February 7, 2008 

Mr. David Staudt 

Center for Disease Control and Prevention 

Acquisition and Assistance Field Branch 

Post Office Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Re: Contract No. 200-2004-03805, Task Order 1: Draft Report SC&A-TR-TASK1-0020, 

Review of the NIOSH Site Profile for the Portsmouth Gaseous Diffusion Plant 

Dear Mr. Staudt: 

SC&A is pleased to submit its draft report, Review of the NIOSH Site Profile for the Portsmouth 

Gaseous Diffusion Plant, SC&A-TR-TASK1-0020. The main body of this report has been 

reviewed for classified information and for Privacy Act (PA) information, edited accordingly, and 

cleared for unrestricted distribution. Attachment 1 of this report, Summary of Site Expert 

Interviews, has not yet been cleared and is not included in this submission. This attachment will be 

forwarded to you at a later date. 

 If you have any comments or questions, please contact me at 732-530-0104. 

Project Manager 

cc: P. Ziemer, PhD, Board Chairperson D. Sundin, NIOSH 

Advisory Board Members A. Makhijani, PhD, SC&A 

L. Wade, PhD, NIOSH H. Behling, PhD, MHP, SC&A 

C. Branche, NIOSH M. Thorne, SC&A 

L. Elliott, NIOSH H. Chmelynski, SC&A 

J. Neton, PhD, NIOSH J. Fitzgerald, Saliant 

S. Hinnefeld, NIOSH J. Lipsztein, SC&A 

L. Homoki-Titus, NIOSH K. Robertson-DeMers, CHP, Saliant 

A. Brand, NIOSH S. Ostrow, PhD, SC&A 

J. Broehm, NIOSH K. Behling, SC&A 

C. Ellison, NIOSH T. Bell, Saliant 

L. Shields, NIOSH Project File (ANIOS/001/20) 

Sincerely, 

John Mauro, PhD, CHP 

1608 SPRING HILL ROAD, SUITE 400 • VIENNA, VIRGINIA • 22182 • 703.893.6600 • FAX 703.821.8236 
 

Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

Review of the NIOSH Site Profile for the Portsmouth Gaseous Diffusion Plant 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0020
 

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road 

Suite 400 

Vienna, Virginia 22182 

Saliant, Inc.
 

5579 Catholic Church Road 

Jefferson, Maryland 21755
 

February 2008 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board 

on Radiation and Worker Health (ABRWH) to maintain all possible openness in its 

deliberations. However, the ABRWH and its contractor, SC&A, caution the reader that at the 

time of its release, this report is pre-decisional and has not been reviewed by the Board for 

factual accuracy or applicability within the requirements of 42 CFR 82. This implies that once 

reviewed by the ABRWH, the Board’s position may differ from the report’s conclusions. Thus, 

the reader should be cautioned that this report is for information only and that premature 

interpretations regarding its conclusions are unwarranted. 
 

 

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S. COHEN & ASSOCIATES: Document No. 

SCA-TR-TASK1-0020 

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

Effective Date: 

Draft — February 6, 2008 

NIOSH Dose Reconstruction Program Revision No. 0 – (Draft) 

REVIEW OF THE NIOSH SITE PROFILE 

FOR THE PORTSMOUTH GASEOUS 

DIFFUSION PLANT 

Page 2 of 89 

Task Manager: 

________________________ Date: ___________ 

U. Hans Behling, PhD, MPH 

Supersedes: 

N/A 

Project Manager: 

________________________ Date: ___________ 

John Mauro, PhD, CHP 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

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TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................5 

1.0 Executive Summary .............................................................................................................7 

1.1 Technical Approach and Review Criteria................................................................7 

1.2 Summary of Findings...............................................................................................8 

1.3 Report Organization ...............................................................................................10 

2.0 Scope and Introduction ......................................................................................................12 

2.1 Review Scope.........................................................................................................12 

2.2 Assessment Criteria and Methods..........................................................................14 

2.2.1 Objective 1: Completeness of Data Sources .............................................14 

2.2.2 Objective 2: Technical Accuracy..............................................................14 

2.2.3 Objective 3: Adequacy of Data.................................................................14 

2.2.4 Objective 4: Consistency among Site Profiles..........................................15 

2.2.5 Objective 5: Regulatory Compliance........................................................15 

3.0 Relevant Background Information.....................................................................................17 

3.1 Principal Operations...............................................................................................17 

3.2 Sources and Quantities of Uranium Hexafluoride Feed Material..........................17 

3.3 Recycled Uranium in Feed Materials ....................................................................18 

3.3.1 Other Sources of Uranium Hexafluoride Feed Materials ..........................18 

3.3.2 Radionuclides of Concern..........................................................................19 

4.0 Findings Identified in Behalf of the Portsmouth Gaseous Diffusion Plant Site Profile ....20 

4.1 Review of TBD-1 (ORAUT-TKBS-0015-1) Portsmouth Gaseous Diffusion 

Plant – Introduction................................................................................................20 

4.2 Review of TBD-2 (ORAUT-TKBS-0015-2) Portsmouth Gaseous Diffusion 

Plant – Site Description .........................................................................................20 

4.3 Review of TBD-3 (ORAUT-TKBS-0015-3) Portsmouth Gaseous Diffusion 

Plant – Occupational Medical Dose.......................................................................21 

4.4 Review of TBD-4 (ORAUT-TKBS-0015-4) Portsmouth Gaseous Diffusion 

Plant – Occupational Environmental Dose ............................................................22 

4.4.1 Internal Environmental Exposures.............................................................22 

4.4.2 External Environmental Dose ....................................................................26 

4.5 Review of TBD-5 (ORAUT-TKBS-0015-5), Portsmouth Gaseous Diffusion 

Plant – Occupational Internal Dose .......................................................................30 

4.6 Review of TBD-6 (ORAUT-TKBS-0015-6) Portsmouth Gaseous Diffusion 

Plant – Occupational External Dose ......................................................................60 

5.0 Worker Interviews, Data Completeness, and Data Integrity .............................................76 

5.1 Relevant Background Information.........................................................................76 

5.2 Site Interviews .......................................................................................................77 

5.3 Document Review..................................................................................................79 

5.4 The Butler 1996 Report .........................................................................................79 

5.5 The Cardarelli (1997) Report.................................................................................81 

5.6 DOE (2000) Report................................................................................................81 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

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5.7 Conclusions Pertaining to Data Integrity and the Adequacy of Data from 

Dose Reconstruction ..............................................................................................83 

6.0 References ..........................................................................................................................85 

Attachment 1: Summary of Site Expert Interviews ......................................................................89 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

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ACRONYMS AND ABBREVIATIONS 

Advisory Board 

or ABWRH Advisory Board on Radiation and Worker Health 

AEC U.S. Atomic Energy Commission 

AGG Argon Gamma Graphs 

ANTI-C’s Anti-contamination Clothing 

BJC Bechtel Jacob Corporation 

Chi/Q Ratio of average air concentration to release rate at the source 

ClF3 Chlorotrifluoride 

CIP Cascade Improvement Program 

CUP Cascade Upgrade Program 

DDE Deep Dose Equivalent 

DOE Department of Energy 

DOELAP Department of Energy Laboratory Accreditation Program 

DOL Department of Labor 

dpm Disintegrations per minute 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

ES&H Environmental Safety and Health 

EWP Electrical Work Permit 

GAT Goodyear Atomic Corporation 

GCEP Gaseous Centrifuge Enrichment Plant 

HASA High Assay Sampling Area 

HEU Highly Enriched Uranium 

HHE Health Hazard Evaluation 

ICN International Chemical and Nuclear Corporation 

K-25 Oak Ridge Gaseous Diffusion Plant 

LEU Low Enriched Uranium 

LOD Limit of Detection 

MDC Minimum Detectable Concentration 

MIVRML Mobile In-Vivo Radiation Monitoring Laboratory 

MRD Minimum Reported Dose 

MTU Metric Ton(ne)s 

NIOSH National Institute for Occupational Safety and Health 

NPDES National Pollutant Discharge Elimination System 

NRC Nuclear Regulatory Commission 

NVLAP National Voluntary Laboratory Accreditation Program 

OCAW Oil, Chemical, and Atomic Workers Union 

OR DOE Oak Ridge Operations Office 

ORAUT Oak Ridge Associated Universities Team 

OW Open Window 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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PAL Plant Allowable Limit 

PGDP Paducah Gaseous Diffusion Plant 

POC Probability of Causation 

PORTS Portsmouth Gaseous Diffusion Plant 

PW Product Withdrawal 

RU Recycled Uranium 

SC&A S. Cohen & Associates 

SDE Shallow Dose Equivalent 

SEC Special Exposure Cohort 

SNM Special Nuclear Material 

SPFPA Security Police and Fire Protection of America 

TBD Technical Basis Document 

TBP Tri-butyl phosphate 

TCE Trichloroethylene 

TIB Technical Information Bulletin 

TLD Thermoluminescent Dosimeter 

TRU Transuranics 

UF4 Uranium Tetrafluoride 

UF6 Uranium Hexafluoride 

U3O8 Uranium Oxide 

UO2F2 Uranyl Fluoride 

UMH Uranium Material Handler 

USEC United States Enrichment Corporation 

USW United Steel Workers 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

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1.0 EXECUTIVE SUMMARY 

This draft report presents the S. Cohen and Associates (SC&A, Inc.) evaluation of the National 

Institute for Occupational Safety and Health (NIOSH) Site Profile for the Portsmouth Gaseous 

Diffusion Plant (PORTS) (ORAUT-TKBS-0015), which was issued as six separate technical 

basis documents (TBDs) numbered ORAUT-TKBS-0015-1 through ORAUT-TKBS-0015-6. 

This draft report was prepared at the request of the Advisory Board on Radiation and Worker 

Health (Advisory Board) and covers all six TBDs identified below. 

•	 ORAUT-TKBS-0015-1, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Introduction, Vol. 1, Rev. 00 (Portsmouth TBD, 2007a) 

•	 ORAUT-TKBS-0015-2, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Site Description, Vol. 2, Rev. 00 (Portsmouth TBD, 2006a) 

•	 ORAUT-TKBS-0015-3, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Medical Dose, Vol. 3, Rev. 00 (Portsmouth TBD, 2006b) 

•	 ORAUT-TKBS-0015-4, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Environmental Dose, Vol. 4, Rev. 01 (Portsmouth TBD, 

2004a) 

•	 ORAUT-TKBS-0015-5, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Internal Dose, Vol. 5, Rev. 00 (Portsmouth TBD, 2004b) 

•	 ORAUT-TKBS-0015-6, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational External Dosimetry, Vol. 6, Rev. 00 (Portsmouth TBD, 

2006c) 

Throughout this report, individual TBDs are referenced simply by number. For example, 

ORAUT-TKBS-0015-1 will be identified as TBD-1. As part of our evaluation, SC&A also 

reviewed numerous other documents that were considered relevant, including the following: 

•	 Select documents that were referenced in the PORTS Site Profile 

•	 Documents contained in the NIOSH Site Research Query Database 

•	 ORAUT-OTIB-0036 – Internal Dosimetry Coworker Data for Portsmouth Gaseous 

Diffusion Plant 

•	 ORAUT-OTIB-0040 – External Dosimetry Coworker Dosimetry Data for the 

Portsmouth Gaseous Diffusion Plant 

1.1 TECHNICAL APPROACH AND REVIEW CRITERIA 

The approach used by SC&A to perform this review follows the procedural protocols described 

in Standard Operating Procedure for Performing Site Profile Reviews (SC&A 2004). Approved 

by the Advisory Board on March 18, 2004, SC&A’s protocol reflects the following review 

criteria: 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

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(1) Completeness of data sources 

(2) Technical accuracy 

(3) Adequacy of data 

(4) Consistency with other site profiles 

(5) Regulatory compliance 

Deficiencies pertaining to these review criteria are noted as 15 “findings,” because these 

represent deficiencies that may require correction due to their potential adverse impact(s) on 

dose reconstruction. The purpose of this review is to provide the Advisory Board with an 

independent assessment of issues that surround the PORTS Site Profile. Findings identified in 

our review are expected to provide the Advisory Board with a preliminary overview of potential 

issues that may impact the feasibility of dose assessment. 

SC&A’s draft report with its preliminary findings will subsequently undergo a multi-step 

resolution process. Resolution includes a transparent review and discussion of draft findings 

with members of the Advisory Board’s Working Group and select personnel representing 

NIOSH/Oak Ridge Associated Universities Team (ORAUT). This resolution process is intended 

to ensure that each finding is evaluated on its technical merit in a fair and impartial manner. A 

final report will then be issued to the full Advisory Board for review and deliberation. 

1.2 SUMMARY OF FINDINGS 

As stated above, SC&A identified a total of 15 findings as a result of our review of the PORTS 

Site Profile. An overview of findings identified in each TBD is presented below. 

TBD-1 (Introduction). The Introduction describes the purpose and scope of the PORTS Site 

Profile. SC&A has no findings regarding information provided in TBD-1. 

TBD-2 (Site Description). The Site Description TBD provides critical information regarding the 

historical and current status of facilities, processes, source terms, etc., at PORTS. SC&A’s 

review identified the following finding: 

 Finding 4.2.1. Based on site experts interviewed by SC&A, TBD-2 failed to 

identify/characterize several buildings/locations at PORTS that had the potential for 

worker exposures. 

TBD-3 (Occupational Medical Dose). The Occupational Medical Dose TBD provides guidance 

for reconstructing doses from diagnostic x-ray procedures required as a condition of 

employment. SC&A’s evaluation of TBD-3 identified the following findings: 

 Finding 4.3.1. At PORTS, available records that identify the use of photofluorography 

are incomplete and do not define a timeframe for its use. Although the TBD identifies 

two timeframes to dose reconstructors (1954–1960 versus 1954–1957), the shorter 

timeframe was selected and was based on a single record with a recorded date of 

October 1957. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

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SC&A views the justification for the shorter time period (i.e., 1954–1957) as technically 

unsound, whimsical at best, and claimant unfavorable. 

TBD-4 (Occupational Environmental Dose). TBD-4 provides data for the reconstruction of 

doses to unmonitored workers exposed onsite internally and externally from environmental 

releases. SC&A’s review identified the following three findings with regard to external 

occupational environmental dose: 

 Finding 4.4-1. Use of the generic ambient environmental dose of 35.9 mrem/y is too 

restrictive for non-compensable claims and claimant unfavorable. 

 Finding 4.4-2. The default ambient environmental dose of 267 mrem/y to workers 

exposed at the Cylinder Storage Yards is without technical basis and may be too low. 

 Finding 4.4-3. Ambient environmental doses are confined to the deep dose that may 

significantly underestimate the potential shallow dose to the skin. 

TBD-5 (Occupational Internal Dose). At PORTS, internal exposure is dominated by uranium 

that existed over a wide range of enrichment. Other radionuclides of concern included 

transuranics and contaminants associated with recycled uranium. 

SC&A’s review of TBD-5 identified the following six findings: 

 Finding 4.5-1. TBD-5 provides activity values for transuranic (TRU) elements and Tc-99 

in reactor tails processed at PORTS. Values cited for Tc-99 were understated by several 

orders of magnitude. 

 Finding 4.5-2. Inconsistent bioassay protocols were employed that significantly affect 

the interpretation of urine bioassay data used for dose reconstruction. 

 Finding 4.5-3. Current guidance for estimating internal exposure to recycled uranium 

(RU) contaminants is unachievable and/or inappropriate. 

 Finding 4.5-4. Empirical data suggest that the generic default value of 3.5% enrichment 

for uranium is inappropriate/claimant unfavorable for large segments of worker 

groupings. 

 Finding 4.5-5. TBD-5 contains contradictory/erroneous data and guidance that instructs 

the use of an incorrect minimum detectable concentration (MDC) value. 

 Finding 4.5-6. Mobile In Vivo Radiation Monitoring Laboratory (MIVRML) chest 

counts for the detection of uranium, TRUs, and fission products are subject to significant 

limitations and uncertainties. 

TBD-6 (Occupational External Dose). At PORTS, radiation fields contributing to external 

radiation included photons, neutrons, and betas. SC&A’s review of TBD-6 identified the 

following four findings: 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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 Finding 4.6-1. The assumed LOD value for shallow dose (as defined by the two-element 

film dosimeter used between 1954 and 1980) lacks technical support and is not claimant 

favorable. 

 Finding 4.6-2. Unmonitored shallow doses derived from coworker data suffer 

deficiencies that are likely the result of dosimeter design limitations and/or process 

policies. 

 Finding 4.6-3. External exposures to localized skin and to extremities were inadequately 

monitored, and guidance to dose reconstructors is too subjective and arbitrary. 

 Finding 4.6-4. Before 1992, PORTS failed to monitor workers for neutron exposures. 

Current guidance to account for unmonitored neutron exposures is incomplete. 

1.3 REPORT ORGANIZATION 

Based on the issues raised in each of the six TBDs, SC&A prepared a summary of findings, 

which is provided above in the executive summary. These findings are not meant to be 

exhaustive, but rather issues of dosimetric significance that SC&A investigated in more detail in 

order to develop suggestions for improvement of any revisions to the PORTS TBD and for use in 

dose reconstruction, as appropriate. 

In accordance with directions provided by the Advisory Board and with site profile review 

procedures prepared by SC&A and approved by the Advisory Board, this report is organized into 

the following sections: 

(1) Executive Summary 

(2) Scope and Introduction 

(3) Relevant Background Information 

(4) Findings Identified on Behalf of the PORTS Site Profile 

(5) Worker Interviews, Data Completeness, and Data Integrity 

Following this Executive Summary, Section 2.0 identifies the review objectives that were used to 

evaluate the PORTS TBD. 

Section 3.0 of this report provides a brief summary of relevant background data contained in the 

PORTS Site Profile. The site profile specifies relevant background information and methods to 

be used by NIOSH for the reconstruction of internal and external doses. Included herein are 

brief summaries of materials and quantities processed, facility descriptions, and radionuclides of 

concern to dose reconstruction. 

As a result of our review of the site profile and other documents, SC&A identified a total of 15 

findings, which are cited in Section 4 of this report. In behalf of each finding, a discussion is 

provided that serves to explain the technical basis for our concern. For some findings, support is 

also provided by one or more documents, which are enclosed as exhibit(s), or are referenced (see 

Reference List in Section 6.0). 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 

 

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These exhibits frequently contain empirical data and/or personal observations/opinions expressed 

by key individuals who were involved in PORTS operations and worker/workplace monitoring. 

As such, SC&A regards these historical documents as highly relevant, credible, and impartial. 

For this reason, the reader is encouraged to review the enclosed exhibits and independently 

determine the degree to which they support each of the corresponding findings. For practical 

reasons, findings are grouped by category in the following subsections of Section 4.0: 

•	 Subsection 4.1: Discusses TBD-1, Introduction, and provides recommendations for 

improvement of this section; however, no findings were identified. 

•	 Subsection 4.2: Findings associated with TBD-2, Site Description. 

•	 Subsection 4.3: Findings associated with the assessment of occupational medical doses 

(TBD-3). 

•	 Subsection 4.4: Findings associated with the assessment of occupational environmental 

doses (TBD-4). 

•	 Subsection 4.5: Findings associated with the assessment of occupational internal doses 

(TBD-5). 

•	 Subsection 4.6: Findings associated with the assessment of occupational external doses 

(TBD-6). 

Section 5.0 discusses worker interviews and presents our opinion(s) regarding data completeness 

and data integrity. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

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2.0 SCOPE AND INTRODUCTION 

Under 42 U.S.C. 7384[14], Congress initially granted the Portsmouth Gaseous Diffusion Plant 

(PORTS) Special Exposure Cohort (SEC) status. Members of the cohort include U.S. 

Department of Energy (DOE) employees, contractors, or subcontractor employees who were 

potentially exposed to radiation and were employed an aggregate of at least 250 workdays before 

February 1, 1992. As provided in 42 U.S.C. 73841(9)[A], members of the cohort who incur one 

(or more) compensable cancers, as specified in Section 3621[17] of the Energy Employees 

Occupational Illness Compensation Program Act of 2000 (EEOICPA) [42 U.S.C. 73841[17], 

qualify for compensation without the need for the U.S. Department of Labor (DOL) to evaluate 

the probability that the cancer was caused by occupational radiation exposures. Excluded from 

SEC compensation are employees with less than 250 days of aggregate employment and 

employees with cancer(s) not specified as compensable. 

In spite of the PORTS SEC statute, NIOSH has determined that it is feasible to reconstruct doses 

for PORTS employees with sufficient accuracy. Correspondingly, NIOSH developed a site 

profile along with two technical information bulletins (TIBs) for PORTS that are to be used for 

dose reconstruction in behalf of claims not covered by SEC criteria. 

2.1 REVIEW SCOPE 

Under the EEOICPA and federal regulations defined in Title 42, Part 82, Methods for Radiation 

Dose Reconstruction Under the Energy Employees Occupational Illness Compensation 

Program, of the Code of Federal Regulations (42 CFR Part 82), the Advisory Board on 

Radiation and Worker Health (Advisory Board) is mandated to conduct an independent review 

of the methods and procedures used by the National Institute for Occupational Safety and Health 

(NIOSH) and its contractors for dose reconstruction. As a contractor to the Advisory Board, S. 

Cohen and Associates (SC&A, Inc.) has been charged under Task 1 to support the Advisory 

Board in this effort by independently evaluating a select number of site profiles that correspond 

to specific facilities at which energy employees worked and were exposed to ionizing radiation. 

This report provides a review of the following six technical basis documents (TBDs) related to 

historical occupational exposures at the Portsmouth Gaseous Diffusion Plant: 

•	 ORAUT-TKBS-0015-1, 2007, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Introduction, Vol. 1, Rev. 00 

•	 ORAUT-TKBS-0015-2, 2006, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Site Description, Vol. 2, Rev. 00 

•	 ORAUT-TKBS-0015-3, 2006, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Medical Dose, Vol. 3, Rev. 00 

•	 ORAUT-TKBS-0015-4, 2004, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Environmental Dose, Vol. 4, Rev. 01 

•	 ORAUT-TKBS-0015-5, 2004, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational Internal Dose, Vol. 5, Rev. 00 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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•	 ORAUT-TKBS-0015-6, 2006, Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant – Occupational External Dosimetry, Vol. 6, Rev. 00 

These documents are referred to in this review as PORTS TBD Volumes 1 through 6. There 

were two TIBs specific to PORTS: (1) Internal Dosimetry Coworker Data for Portsmouth 

Gaseous Diffusion Plant (ORAUT-OTIB-0036, Rev. 00) and (2) External Coworker Dosimetry 

Data for Portsmouth Gaseous Diffusion Plant (ORAUT-OTIB-0040, Rev. 00). SC&A also 

reviewed other pertinent documents, including those cited on the NIOSH Site Research database. 

SC&A, in support of the Advisory Board, has critically reviewed the PORTS TBDs, as well as 

supplementary and supporting documents, against the following three evaluation criteria: 

•	 Determine the completeness of the information gathered by NIOSH, with a view to 

assessing its adequacy and accuracy in supporting individual dose reconstructions 

•	 Assess the technical merit of the data/information 

•	 Assess NIOSH’s guidelines for the use of the data in dose reconstructions 

SC&A’s review of the six volumes that comprise the TBD, along with its supporting 

supplemental documentation, focuses on the quality and completeness of the data that 

characterized the facility and its operations, and the adequacy of these data in dose 

reconstruction. The review was conducted in accordance with SC&A Standard Operating 

Procedure for Performing Site Profile Reviews (SC&A 2004), which was approved by the 

Advisory Board. 

The review is directed at “sampling” the site profile analyses and data for validation purposes. 

The review does not provide a rigorous quality control process, whereby actual analyses and 

calculations are duplicated or verified. The scope and depth of the review are focused on aspects 

or parameters of the site profile that would be particularly influential in dose reconstructions, 

bridging uncertainties, or correcting technical inaccuracies. This review does not explicitly 

address the issue of radiation exposures to cleanup workers and decommissioning workers, as 

that is not addressed in the TBDs. 

The six volumes of the PORTS Site Profile are supposed to serve as site-specific guidance 

documents to be used in support of dose reconstructions. While dose reconstructors use other 

data, information, and guidance documents in making dose estimates, the purpose of site profiles 

is to provide dose reconstructors with consistent general information and specifications to 

support their individual dose reconstructions. This report was prepared by SC&A to provide the 

Advisory Board with an evaluation of whether and how the TBDs can support the various types 

of dose reconstruction estimates that NIOSH performs—minimum for compensation only; 

maximum, with worst-case assumptions to be used for denial only, and “best-case” or 

“reasonable” dose estimates to be used for both compensation and denial. The criteria for 

evaluation include whether the TBDs provide a basis for scientifically supportable and claimantfavorable dose reconstructions that systematically resolve uncertainties in favor of the claimant 

as required by 42 CFR 82, the regulation governing the dose reconstruction process. 

The basic principle of dose reconstruction is to characterize the radiation environments to which 

workers were exposed, and determine the levels of exposure the workers received in those 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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environments through time. The hierarchy of data used for developing dose reconstruction 

methodologies is dosimeter readings and bioassay data, coworker data and workplace monitoring 

data, and process description information or source term data. 

SC&A’s review of the PORTS TBDs was further supplemented by interviews with site 

personnel in order to gain a better insight into operational practices and the implementation of 

radiation protection protocols. Attachment 1 provides a transcript of the interviews, in which 

statements were paraphrased and names of those interviewed have been omitted for privacy 

reasons. While discussions with site personnel avoided issues that to date may be regarded as 

classified, the interviews (as summarized in Attachment 1) were submitted to the DOE for 

review as a precautionary measure. 

2.2 ASSESSMENT CRITERIA AND METHODS 

Under Task 1, SC&A is charged with evaluating the approach set forth in the site profiles that is 

used in the individual dose reconstruction process. These documents are reviewed for their 

completeness, technical accuracy, adequacy of data, consistency with other site profiles, and 

compliance with the stated objectives, as defined in SC&A Standard Operating Procedure for 

Performing Site Profile Reviews (SC&A 2004). This review is specific to the PORTS Site 

Profile and supporting TIBs. Our review identifies a number of issues, and discusses the degree 

to which the site profile fulfills the review objectives delineated in SC&A 2004 in behalf of the 

following objectives. 

2.2.1 Objective 1: Completeness of Data Sources 

SC&A reviewed the site profile with respect to Objective 1, which requires SC&A to identify 

principal sources of data and information that are applicable to the development of the site 

profile. The two elements examined under this objective include (1) determining if the site 

profile made use of available data considered relevant and significant to the dose reconstruction, 

and (2) investigating whether other relevant/significant sources are available, but were not used 

in the development of the site profile. 

2.2.2 Objective 2: Technical Accuracy 

Objective 2 requires SC&A to perform a critical assessment of the methods used in the site 

profile to develop technically defensible guidance or instructions, including evaluating field 

characterization data, source term data, technical reports, standards and guidance documents, and 

literature related to processes that occurred at PORTS. The goal of this objective is to analyze 

the data according to sound scientific principles, and then to evaluate this information in the 

context of dose reconstruction. 

2.2.3 Objective 3: Adequacy of Data 

Objective 3 requires SC&A to determine whether the data and guidance presented in the site 

profile are sufficiently detailed and complete to conduct dose reconstruction, and whether a 

defensible approach has been developed in the absence of data. In addition, this objective 

requires SC&A to assess the credibility of the data used for dose reconstruction. The adequacy 

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of the data identifies gaps in the facility data that may influence the outcome of the dose 

reconstruction process. For example, if a site did not monitor all workers exposed to neutrons 

who should have been monitored, this would be considered a gap and thus an inadequacy in the 

data. An important consideration in this aspect of our review of the site profile is the scientific 

validity and claimant favorability of the data, methods, and assumptions employed in the site 

profile to fill in data gaps. 

2.2.4 Objective 4: Consistency among Site Profiles 

Objective 4 requires SC&A to identify common elements within site profiles completed or 

reviewed to date, as appropriate. In order to accomplish this objective, the PORTS TBDs were 

compared to other TBDs reviewed to date. This assessment was conducted to identify areas of 

inconsistencies and determine the potential significance of any inconsistencies with regard to the 

dose reconstruction process. 

2.2.5 Objective 5: Regulatory Compliance 

Objective 5 requires SC&A to evaluate the degree to which the site profile complies with stated 

policy and directives contained in 42 CFR Part 82. In addition, SC&A evaluated the TBD for 

adherence to general quality assurance policies and procedures utilized for the performance of 

dose reconstructions. 

In order to place the above objectives into the proper context as they pertain to the site profile, it 

is important to briefly review key elements of the dose reconstruction process, as specified in 

42 CFR Part 82. Federal regulations specify that a dose reconstruction can be broadly placed 

into one of three discrete categories. These three categories differ greatly in terms of their 

dependence on and the completeness of available dose data, as well as on the 

accuracy/uncertainty of data. 

Category 1: Least challenged by any deficiencies in available dose/monitoring data are dose 

reconstructions for which even a partial assessment [or minimized dose(s)] corresponds to a 

probability of causation (POC) value in excess of 50%, assuring compensability to the claimant. 

In some cases, such partial/incomplete dose reconstructions with a POC greater than 50% may 

involve only a limited amount of external or internal data. In extreme cases, even a total absence 

of a positive measurement may suffice for an assigned organ dose [based on limits of detection 

(LOD)] that results in a POC greater than 50%. For this reason, dose reconstructions in behalf of 

this category may only be marginally affected by incomplete/missing data or uncertainty of the 

measurements. In fact, regulatory guidelines recommend the use of a partial/incomplete dose 

reconstruction, the minimization of dose, and the exclusion of uncertainty for reasons of process 

efficiency, as long as this limited effort produces a POC of greater than or equal to 50%. 

Category 2: A second category of dose reconstruction defined by federal guidance recommends 

the use of “worst-case” assumptions. The purpose of worst-case assumptions in dose 

reconstruction is to derive maximal or highly improbable dose assignments. For example, a 

worst-case assumption may place a worker at a given work location 24 hours per day and 

365 days per year. The use of such maximized (or upper bound) values, however, is limited to 

those instances where the resultant maximized doses yield POC values below 50%, which are 

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not compensated. For this second category, the dose reconstructor needs only to ensure that all 

potential internal and external exposure pathways have been considered, and that the approach is 

scientifically supportable. 

The obvious benefit of worst-case assumptions and the use of maximized doses in dose 

reconstruction is efficiency. Efficiency is achieved by the fact that maximized doses avoid the 

need for precise data and eliminates consideration for the uncertainty of the dose. Lastly, the use 

of bounding values in dose reconstruction minimizes any controversy regarding the decision not 

to compensate a claim. 

Although simplistic in design, the TBD must, at a minimum, provide information and data that 

clearly identify (1) all potential radionuclides, (2) all potential modes of exposure, and (3) upper 

limits for each contaminant and mode of exposure to satisfy this type of a dose reconstruction. 

Thus, for external exposures, maximum dose rates must be identified in time and space that 

correspond to a worker’s employment period, work locations, and job assignment. Similarly, in 

order to maximize internal exposures, highest air concentrations and surface contaminations 

must be identified. 

Category 3: The most complex and challenging dose reconstructions consist of claims where 

the case cannot be dealt with under one of the two categories above. For instance, when a 

minimum dose estimate does not result in compensation, a next step is required to make a more 

complete estimate. Or when a worst-case dose estimate that has assumptions that may be 

physically implausible results in a POC greater than 50%, a more refined analysis is required. A 

more refined estimate may be required either to deny or to compensate. In such dose 

reconstructions, which may be represented as a “reasonable” or “best-case” estimate, NIOSH has 

committed to resolve uncertainties in favor of the claimant. According to 42 CFR 82, NIOSH 

interprets “reasonable estimates” of radiation dose to mean the following: 

... estimates calculated using a substantial basis of fact and the application of 

science-based, logical assumptions to supplement or interpret the factual basis. 

Claimants will in no case be harmed by any level of uncertainty involved in 

their claims, since assumptions applied by NIOSH will consistently give the 

benefit of the doubt to claimants. [Emphasis added.] 

SC&A’s draft report and preliminary findings will subsequently undergo a multi-step resolution 

process. Resolution includes a transparent review and discussion of draft findings with members 

of the Advisory Board Working Group, petitioners, claimants, and interested members of the 

public. This resolution process is intended to ensure that each finding is evaluated on its 

technical basis in a fair and impartial basis. A final report will then be issued to the full Board 

for deliberation and a final recommendation. 

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3.0 RELEVANT BACKGROUND INFORMATION 

This section presents summary information that will provide the reader with an overview of key 

facility processes, production quantities, and radiological source terms that may have contributed 

to internal and external exposures. 

3.1	 PRINCIPAL OPERATIONS 

The PORTS began operation in 1954. For 47 years (i.e., until 2001), the primary purpose of 

PORTS was the enrichment of uranium. Enrichment by gaseous diffusion involved uranium 

hexafluoride (UF6) gas as feed material, which is passed through a series of semiporous barriers 

or stages. Each stage allows the lighter U-235 to pass through more easily than the heavier U238. This results in each successive stage having slightly more U-235 on one side and, therefore, 

increasing (i.e., enriches) the amount of fissile U-235. The enriched UF6 gas flows up towards 

the top of the cascade, while depleted UF6 gas remains at the bottom portion of the stage. 

In total, the PORTS cascade consisted of 4,080 stages that were housed in three separate process 

buildings. The initial enrichment of UF6 was performed in Building X-333, followed by 

intermediate enrichment in Building X-330. Process Building X-326 contained the highest stage 

enrichment cascade and was able to enrich uranium to more than 97%. 

At each stage, variably enriched as well as depleted UF6 was shipped offsite for further 

processing and use in military applications, and lower enriched UF6 was converted to 

commercial reactor fuel. Depleted UF6 (or tails) were either re-fed to the cascade or stored 

onsite. 

Because UF6 is a solid at ambient temperature, it is received and/or shipped as a solid in various 

cylinder types and sizes, depending on the level of enrichment and associated concerns for 

nuclear criticality. 

3.2	 SOURCES AND QUANTITIES OF URANIUM HEXAFLUORIDE FEED 

MATERIAL 

During the 47 years of uranium enrichment at PORTS, about 330,000 Metric Ton(ne)s uranium 

(MTU) in the form of UF6 passed through the enrichment cascades. Included were an estimated 

121,485 MTU of enriched UF6 withdrawn from the Paducah Gaseous Diffusion Plant (PGDP) 

cascades and supplied to PORTS as feed for further enrichment. A second major source of feed 

material came from the K-25 Plant at Oak Ridge. 

Only about 11,890 MTU of UF6 feed was produced at PORTS between 1958 and 1962 at the 

Feed Manufacturing Plant, Building X-344. This facility converted UF4 (green salt provided by 

Mallinckrodt Chemical Works and National Lead of Ohio) to UF6 in fluorination towers, where 

powdered green salt passed through a rising column of fluorine gas. 

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3.3 RECYCLED URANIUM IN FEED MATERIALS 

Among the UF6 feed materials processed at PORTS were an estimated 1,094.66 MTU of 

recycled uranium (RU). All but 1.86 MTU of RU-containing feed had been provided by PGDP 

and the Oak Ridge K-25 Plant between 1955 and 1974. This reactor return feed material 

consisted of slightly depleted uranium with a U-235 content between 0.64% and 0.68%. 

3.3.1 Other Sources of Uranium Hexafluoride Feed Materials 

Buildings X-705 and X-705E. Other feed material was produced by the Decontamination, 

Cleaning, and Recovery (Building X-705) and the Oxide Conversion Facility (Building X-705E). 

At the uranium recovery facility (Building X-705), uranium-bearing waste streams and uranium 

scrap were processed to extract uranium in the form of U3O8. Between 1958 and 2001, this 

facility recovered 38.2 MTU of U3O8. 

The U3O8 produced in Building X-705 was converted directly to UF6 in Building X-705E. This 

facility operated between 1957 and 1978, and produced a total of 233 MTU of UF6. 

Building X-344 (Feed Manufacturing Plant). Uranium tetrafluoride (UF4) produced at 

Mallinckrodt and Fernald was converted in fluorination towers to UF6. Between 1958 and 1962, 

this facility produced 11,890 MTU of UF6 feed material that was processed at PORTS. 

Other Support Facilities: 

•	 Building X-342 (Fluorine Generation Facility). This facility produced fluorine gas for 

the uranium recovery operations in Building X-705E, and housed equipment needed to 

sample and heat UF6 to a gaseous state before feeding to the cascades. 

•	 Building X-343 (Fixed Feed Facility). Initially, UF6 feed and processed material could 

enter or be withdrawn from the cascades at any location. In later years, feed and 

withdrawal were reduced to limited points along the cascades. These fixed feed facilities 

were housed in Building X-343. 

•	 Building X-345 (Special Nuclear Materials Storage). Select areas of Building X-345 

contained vaults that were used to store highly enriched uranium (HEU). This facility 

also contained a high assay sampling area (HASA) for assaying the highly enriched 

UF6. 

•	 Building X-710 (Analytical and Process and Materials Technology Labs). This facility 

calibrated radiological instruments, performed industrial radiograph, and evaluated feed 

and other process materials. 

•	 Building X-744 (Aluminum Smelter and Recovery). Retired process equipment, 

including cascade compressors with residual contamination, was subject to a smelter 

process that extracted aluminum for reuse/resale. 

•	 Building X-770 (Test Loop Facility). This facility was involved in the development/ 

testing of various cascade components (e.g., piping, pumps, sampling lines, etc.). 

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3.3.2 Radionuclides of Concern 

Workers at PORTS were engaged in many process operations and maintenance activities that 

had the potential for both external and internal exposures to a host of radionuclides, as defined in 

Table 1. A credible assessment of worker exposures to these radionuclides is hampered as a 

result of the following: 

•	 Many radionuclides existed in various chemical states that impacted their physical and 

biological behavior in the environment and in the human body. 

•	 Chemical processes altered the relative abundance of individual radionuclides by 

selectively concentrating some radionuclides in finished products, while concentrating 

others in waste streams and tailings. 

•	 Although isotopes of uranium were the dominant concern, feed materials at PORTS 

contained significant quantities of uranium decay products. 

•	 Among feed materials processed at PORTS was ~1,095 MTU of RU containing 

transuranics and the fission product Tc-99. An estimated total of 60,000 to 90,000 grams 

of Tc-99 was fed into the cascade. 

•	 For the first 40 years of PORTS operation, routine in-vitro bioassays (fluorophotometry 

and gross alpha counting of urine samples) were limited to uranium. Only as recently as 

1994, urinalysis assessed the excretion of Np-237, Pu-238/-239/-240, and Am-241. 

Starting in 1965, the inhalation of uranium compounds was assessed by means of chest 

counting. 

Table 1. List of Radionuclides of Concern 

Radionuclides of Concern 

U-234 

U-235 

U-236 

U-238 

Tc-99 

Np-237
 

Pu-238c 

Pu-239c 

Pu-240c 

Am-241c 

Th-228 

Th-230 

Th-231 

Th-232 

Th-234d 

Pa-234md
 

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4.0 FINDINGS IDENTIFIED IN BEHALF OF THE PORTSMOUTH 

GASEOUS DIFFUSION PLANT SITE PROFILE 

This section identifies findings that resulted from our review of the six TBDs that represent the 

PORTS Site Profile. Findings are grouped by their corresponding TBD and follow the order in 

which the TBDs are sequenced in the Site Profile. For some findings, supportive information is 

provided by one or more exhibits. For ease of tracking, findings and their associated exhibits are 

numbered in a manner that provides a linkage. For example, all findings associated with internal 

dose reflect TBD-5 (of ORAUT-TKBS-0015-5) and are discussed in Section 4.5 below. The 

first finding pertaining to internal dose is, therefore, identified as Finding 4.5-1. There are two 

exhibits that support Finding 4.5-1, which are further labeled as Exhibit 4.5-1A and 4.5-1B. 

4.1	 REVIEW OF TBD-1 (ORAUT-TKBS-0015-1) PORTSMOUTH GASEOUS 

DIFFUSION PLANT – INTRODUCTION 

The Introduction explains the purpose and the scope of the site profile. SC&A was attentive to 

this section, because it provides a useful overview and explains the role of each TBD in support 

of the dose reconstruction process. Hence, the introduction helps in framing the scope of the site 

profile. As will be discussed later in this report, NIOSH may elect to include additional 

qualifying information in the introduction describing the dose reconstruction issues that are not 

explicitly addressed in the TBDs that follow. NIOSH may also want to include a roadmap to the 

dose reconstructor expounding on the entire dose reconstruction process and showing how the 

TBDs fit into the process. 

In spite of its brevity, SC&A identified no findings in behalf of TBD-1. 

4.2	 REVIEW OF TBD-2 (ORAUT-TKBS-0015-2) PORTSMOUTH GASEOUS 

DIFFUSION PLANT – SITE DESCRIPTION 

The Site Description is an important document, because it provides a description of the facilities 

and processes, as well as historical information that serve as the underpinning for subsequent 

TBDs. Specifically, this document describes the history and current status of key facilities and 

processes, and the associated source terms that are relevant to dose reconstruction. SC&A’s 

review of this section focused on whether all the potentially important site activities and 

processes are described, and whether characterization of source terms is complete and sufficient 

to support dose reconstruction. 

It should be noted that much of the relevant background information contained in Section 3.0 of 

this report was taken from TBD-2. A positive feature of TBD-2 is a series of tables (i.e., Table 

2-1 through Table 2-16) that provide a comprehensive overview of key facilities, material 

quantities, and radionuclides of concern. However, based on site experts interviewed by SC&A, 

several facilities/locations at PORTS with the potential for exposures were not included. 

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Finding 4.2-1: TBD-2 (PORTS Site Description) may not be Complete with regard to 

Buildings/Locations with the Potential for Worker Exposures 

Site experts, with employment as far back as 1954, informed SC&A that the current TBD-2 

(ORAUT-TKBS-0015-2) is incomplete with regard to buildings/locations where workers may 

have received radiation exposures. The following were identified by site experts: 

Building Description Comments 

X-746 Cylinder Storage Incident occurred in southeast corner 

X-760 Test loop from the 710 Laboratory 

X-744H Storage Storage of potentially contaminated 

components 

X-744J Storage Storage of potentially contaminated 

components 

X-745G Empty Cylinder Storage Yard 

X-745C Empty Cylinder Storage Yard 

X-745F Heel Cylinder Yard 

The extent to which this deficiency may impact dose reconstruction for select workers who may 

have been assigned to these locations depends on whether these workers were either monitored 

and/or qualify for inclusion in the coworker models, as defined in ORAUT-OTIB-0036 and 

ORAUT-OTIB-0040. 

4.3	 REVIEW OF TBD-3 (ORAUT-TKBS-0015-3) PORTSMOUTH GASEOUS 

DIFFUSION PLANT – OCCUPATIONAL MEDICAL DOSE 

Between 1954 and 1989, annual chest x-rays were mandatory for all employees. Starting in 

1990, the frequency of chest x-rays was reduced to every 3, 5, or 10 years, depending on age. 

Moreover, these diagnostic exams were optional with the exception of asbestos/beryllium 

workers, for whom an annual chest x-ray remained mandatory. Because records pertaining to xray equipment and operating parameters are scarce, generic values cited in ICRP Publication 34 

(ICRP 1983) were used to derive organ-specific doses for conventional 14 in × 17 in posterioranterior and lateral chest x-rays for the years 1961 and beyond. For the earlier years, the absence 

of data raises the potential that photofluorography may have been used in lieu of conventional 

chest x-rays. To account for the potential use of photofluorography, higher organ doses were 

derived that were based on a generic entrance kerma of 3 rads for the years 1954 to 1957. 

In the absence of site-specific data, the use of generic/surrogate data defined in the ICRP and/or 

NCRP publication is not unreasonable for the assignment of organ doses, as given in Tables A-1, 

A-2, A-3 and A-4 of ORAUT-TKBS-0015-3. However, there exists an inconsistency for the 

timeframe for the potential use of photofluorography, as noted in Finding 4.3-1 below. 

Finding 4.3-1: Two Different Timeframes for Assigning Organ Doses Derived for 

Photofluorography Exams are Cited in the TBD 

Section 3.3.1 of TBD-3 states the following: 

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Even-though no evidence of the use of photofluorography was found at PORTS it 

is reasonable to presume that at least some of the occupational medical 

diagnostic chest x-rays with the DOE and its predecessor organizations were 

accomplished by photofluorography. The use of photofluorography should be 

assumed to ensure claimant-favorable dose reconstructions from the time-period 

of 1954 through 1960. [Emphasis added.] 

Section 3.4.1, entitled Photofluorography (1954–1957), identifies a more restrictive time period 

and cites the following justification: 

. . . The PFG period of applicability is based upon the discovery of one PFG 

performed as appeared in one claimant’s file in October of 1957. 

The restrictive time period of 1954 to 1957 for PFG is selected for organ doses defined in 

Tables A-1 and A-2 of the TBD. 

SC&A not only regards the above-cited statements inconsistent, but views the justification of 

restricting the time period of 1957 whimsical at best and claimant unfavorable. 

4.4	 REVIEW OF TBD-4 (ORAUT-TKBS-0015-4) PORTSMOUTH GASEOUS 

DIFFUSION PLANT – OCCUPATIONAL ENVIRONMENTAL DOSE 

The application of data cited in TBD-4 is intended for use in dose reconstruction for 

unmonitored workers. Derived estimates are defined for internal and external exposures. 

PORTS has estimated that approximately 10,545 kg of uranium corresponding to 8 Ci and 27 Ci 

of Tc-99 have been released to the atmosphere from 1955 to 1993. Nearly half (or about 4,800 

kg) of the estimated release of uranium was attributable to one accidental release from a 14-ton 

cylinder in 1978, and another 3,250 kg of the total uranium were released in the first 8 years of 

plant operation when the Feed Production Plant was operational. After the shutdown of the Feed 

Production Plant in 1962, routine environmental releases of uranium dropped significantly 

(Rumble 1978). 

Although Tc-99 was introduced into the cascade system along with feed produced from recycled 

uranium, it was not until 1975 that its presence was first noted. Most of the estimated 27 Ci of 

Tc-99 was likely released during maintenance of contaminated cascade equipment (i.e., Cascade 

Improvement Program/Cascade Upgrade Program or CIP/CUP). 

4.4.1	 Internal Environmental Exposures 

The TBD acknowledges the fact that environmental monitoring only began in 1964, and for most 

years thereafter, focused on monitoring locations that were either offsite or near the PORTS’ 

perimeter. For these reasons, upper-bound estimates of potential intakes of airborne 

contaminants were based on a model that employed annual release quantities and empirical 

yearly Chi/Q values. For years prior to 1964, the maximum of all empirical Chi/Q values was 

used to estimate airborne levels and intakes based on a yearly inhalation of 2,400 m3 . Table 

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4.2.5-1 of the TBD provides maximum annual intakes and their uncertainties for uranium 

isotopes, uranium daughters, and Tc-99. Environmental releases of Np and Pu were considered 

insignificant and were not included. 

As part of SC&A’s evaluation, we compared the modeled intake data defined in Table 4.2.5-1 

against reported environmental air concentrations for select years. For example, for the years 

1964 and 1968, Table 4.2.5-1 of TBD-4 cites total uranium intakes of 17.76 Bq/yr and 8.88 

Bq/yr, respectively. Empirical air sampling data for these 2 years are provided in GAT-449 

(GAT 1964) and GAT-R-568 (GAT 1969). 

Exhibits 4.4-1A and 4.4-1B contain the reported air sampling data for the years 1964 and 1968 at 

various locations. Thus, air sampling data for the year 1964 as given in Exhibit 4.4-1A shows an 

average air concentration of 0.01 × 10-11 

μCi/cc (or 3.7 Bq/m3) for a total of 28 sample locations. 

For the inhalation value of 2,400 m3/y, this would correspond to 8.8 Bq/y, as cited in Table 

4.2.5-1 of TBD-4. 

For 1968, Exhibit 4.4-1B shows that the Phone Building was the location with the highest 

airborne alpha activities, yielding a yearly average of 1.334 × 10-3 dpmα/ft3

 (or 0.047 dpm/m3). 

For the inhalation volume of 2,400 m3

/y, the inhalation intake at the Phone Building is 

estimated at 113 dpm/y or 1.88 Bq/y. 

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EXHIBIT 4.4-1A 

(Source: GAT 1964) 

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EXHIBIT 4.4-1B 

(Source: GAT 1969) 

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The empirically derived alpha intake of 1.88 Bq/y is nearly 5 times lower than the maximized 

modeled value of 8.88 Bq/y proposed by NIOSH. Although SC&A recognizes that the empirical 

value of 1.88 Bq/y represents an offsite location, it is nevertheless concluded that the modeled 

doses defined in Table 4.2.5-1 are likely to represent reasonable, if not maximized, intake values. 

SC&A has no findings pertaining to occupational environmental internal dose. 

4.4.2 External Environmental Dose 

Between 1954 and 1981, onsite ambient dose rates were measured periodically with an openwindow GM tube. Due to instrument limitations, these measurements are of questionable value 

for dose reconstruction. Starting in 1981, onsite environmental gamma radiation levels were 

monitored by thermoluminescent dosimeters (TLDs) at select outdoor locations proximal to 

Perimeter Road that circumscribes PORTS buildings/facilities. Table 4.3.1-1 of TBD-4 provides 

yearly, location-specific onsite doses that are based on a 2,000 hr/y exposure. For all years and 

all locations, the highest annual environmental dose of 35.9 mrem is cited for 1993 at location 3 

on the south side of PORTS near the south holding pond. However, in order to simplify matters, 

Section 4.3.1.1 provides the following guidance to dose reconstructions: 

If background is not to be subtracted, the maximum value of 35.9 mrem from 

Table 4.3.1-1 can be used to assign annual ambient environmental dose to 

workers in areas near Perimeter Road such as the general employee parking lots; 

the guard gates on the outer perimeter of the security area; the switchyards; 

warehouses X-744S, T, and U; process buildings, and wastewater facility X-611. 

(The values in Table 4.3.1-1 for location 874 are much higher, but are specific to 

the UF6 cylinder storage yards and not to the rest of the facility.) If background 

is to be subtracted, the ambient radiation for these areas should be assigned a 

value of 0 mrem because of the argument presented above. [Emphasis added.] 

For location 874, which includes the Cylinder Storage Yards, guidance includes the following: 

Given the above information, an ambient radiation dose of 267 mrem (2,000 hour 

work-year) should be applied to cylinder yard workers as claimant favorable. 

The portion of this dose attributable to neutrons can be assumed to be 178 mrem 

per year. This value for the annual neutron dose is obtained from Table 4.3-3 as 

the maximum annual deep dose equivalent reported in the period 1998 to 2001 

for the cylinder storage yards X-745C and X-745E. This assumes that the 

neutron dose is the only contributor to the deep dose equivalent. All of these 

assumptions are claimant favorable. [Emphasis added.] 

For convenience, Table 4.3-3 and Table 4.3.1-1 of TBD-4 are reproduced herein as Exhibits 4.4­

2A and 4.4-2B, respectively. 

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EXHIBIT 4.4-2A 

Designation 

(Other designations used over the years for 

measurements taken in this vicinity) 

Location 

X-7725 Waste Storage Facility. 

X-326 Process Building. 

X-330 Process Building. 

X-333 Process Building. 

X-344 Containment Building 

X-345 Special Nuclear Material (SNM) Storage Building. 

X-705 Decontamination Building. 

X-720 Maintenance and Stores Building. 

X-744G Bulk Storage Building. 

X-745C, X-745E Depleted Uranium Cylinder Storage Yards. 

Site Interior Ambient Radiological Conditions 

Mean annual dose by location(a), (b) 

(mrem/year) 

Year 

X-7725 X-326 X-345 X-744G X-745C X-745E 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

Avg. 

deep 

dose 

Avg. 

shallow 

dose 

1998 21 36 2 4 19 26 — — 35 39 20 21 

1999 5 10 1 4 1 3 23 19 37 30 56 47 

2000 14 — 0 — 0 — 285 — 122 — 178 — 

2001 23 — 0 — 0 — 1,056 — 142 — 175 — 

a. 2,000-hour work-year, prorated from an 8,736-hour year. 

b. Includes beta, gamma, and neutron. Neutrons included in deep dose. — No data. 

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EXHIBIT 4.4-2B 

Mean annual dose by location(a), (b) 

(mrem/year) 

Year 874 PP518 29 

(10, PP862) 

24 

(PP906) 

12 

(PP933) PP1406 3 (39, X-230-J2, A35) 

A40 

(35, (X-104) 

A36 

(X-611) 

A39 

(PP722) 

1954 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8 

1972 — — 23.6 23.8 22.4 — 24.0 — — — 

1973 — — 27.8 26.4 25.6 — 27.2 — — — 

1974 — — 28.2 27.0 26.0 — 26.8 — — — 

1975 — — 19.2 18.8 18.4 — 18.8 — — — 

1976 — — 20.6 20.4 20.0 — 20.0 — — — 

1977 — — 23.8 24.0 24.2 — 24.4 — — — 

1978 — — 21.8 21.4 20.6 — 21.2 — — — 

1979 — — 21.8 21.2 22.0 — 20.2 — — — 

1980 — — 21.4 20.8 20.8 — 21.6 — — — 

1981 — — 21.0 22.6 20.2 — 19.8 — — — 

1982 — — 14.4 15.6 14.6 — 16.2 17.4 — — 

1983 — — 15.6 12.8 13.4 — 15.2 15.8 — — 

1984 — — 12.4 6.4 12.8 — 20.4 11.4 — — 

1985 — — 13.8 13.8 13.8 — 13.8 13.8 — — 

1986 239.3 — 20.3 16.4 19.1 — 16.9 16.4 20.6 19.1 

1987 267.5 — 21.6 20.0 21.4 — 18.9 19.4 24.5 22.8 

1988 260.6 — 19.6 17.4 19.4 — 20.8 16.6 23.2 21.6 

1989 257.2 — 22.7 20.8 20.7 — 20.7 17.4 22.5 22.7 

1990 250.7 — 23.7 17.9 24.7 — 20.0 19.6 21.0 — 

1991 261.0 16.7 21.5 17.0 21.7 21.7 20.1 NR 20.9 — 

1992 246.5 29.1 36.6 17.4 19.9 44.8 33.7 NR 25.6 — 

1993 150.4 24.8 25.4 24.0 23.6 12.2 35.9 26.0 26.9 — 

1994 27.8 22.9 24.6 30.3 24.6 22.1 29.8 19.9 32.3 — 

1995 90.0 20.0 16.3 16.8 17.4 17.3 18.1 15.5 19.4 — 

1996 112.0 14.0 22.1 26.9 28.4 13.4 15.7 10.5 29.7 — 

1997 131.4 13.8 21.5 25.9 27.5 13.9 15.2 10.4 28.5 — 

1998 135.0 22.0 23.3 23.3 24.3 23.1 24.7 20.4 24.3 — 

1999 131.4 18.5 19.2 16.9 20.6 20.8 20.4 15.8 20.1 — 

2000 149.2 24.5 22.0 20.6 29.5 22.0 24.0 19.7 23.8 — 

2001 150.4 19.1 23.2 16.8 27.2 19.2 20.8 15.7 19.5 — 

2002 142.4 18.4 23.3 15.7 27.1 18.8 20.4 14.8 18.4 — 

a. 2,000-hour work-year, prorated from an 8,736-hour year. 

b. Data for all years are from PORTS annual environmental reports with the exception of 1993 – 2002 for locations
 

PP518, PP862,
 

PP906, PP933, PP1406, A35, A40, A36 and 874, which were obtained from USEC by separate communication (ATL 

2003f). 

— No data. 

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Finding 4.4.2-1: Use of the Generic Ambient Environmental Dose of 35.9 mrem/y is too 

restrictive for non-compensable claims 

SC&A interprets the above-cited statements/guidance to mean that the generic ambient 

environmental dose of 35.9 mrem/y is to apply for non-compensable claims. We infer this from 

the statement that “...if background is to be subtracted, the ambient radiation for these areas 

should be assigned a value of 0 mrem because of the argument presented above.” 

Implicit in this statement is the assumption that there were no significant environmental releases 

at PORTS that would give rise to radiation dose rates above natural background. SC&A 

disagrees with this implicit assumption. 

Finding 4.4.2-2: The Default Ambient Environmental Dose of 267 mrem/y to Workers 

Exposed at the Cylinder Storage Yards (i.e., location 874) is without Technical Basis and 

may be Too Low 

Table 4.3.1-1 of TBD-4 (see Exhibit 4.4-2B) identifies the origin of the 267 mrem/y value as that 

corresponding to location 874 for the year 1989; and Table 4.3-3 of TBD-4 identifies the value 

178 mrem/y as the average deep dose for location X-745E (depleted uranium cylinder storage 

yard) for the year 2000. 

The difficulty in mating these two independent measurements is due to the following facts: 

•	 TLD measurements in 1989 and in 2000 represent two very different time 

periods. 

•	 TLD measurements in 1987 were recorded using Department of Energy 

Laboratory Accreditation Program (DOELAP) TLDs, which were neither 

calibrated nor processed for neutron exposure. To what extent the annual dose of 

267 mrem incorporates a neutron dose component is therefore unknown. In the 

absence of data that might resolve this question, a more favorable approach would 

combine the 267 mrem dose of 1987 with the 178 mrem neutron dose of 2000. 

•	 Section 4.3.1.2 of TBD-4 states that: 

. . . This value [i.e., 178 mrem/y] for the annual neutron dose is obtained from 

Table 4.3-3 as the maximum annual deep dose equivalent reported in the 

period 1998 to 2001 for the cylinder storage yards X-745C and X-745E. 

[Emphasis added] 

Inspection of Exhibit 4.4-2A (i.e., Table 4.3-3 of TBD-4), however, identifies the much larger 

annual dose of 1,956 mrem/y for X-744G. TBD-4 provides no explanation why this dose was 

not considered. 

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Finding 4.4.2-3: Ambient Environmental Doses are Confined to the Deep Dose that may 

Significantly Underestimate Potential Exposures to the Skin 

Claims involving skin cancer as well as other surficial tissues must be evaluated on the basis of 

shallow dose estimates. Dose estimates involving external ambient environmental doses are 

restricted to the deep dose, which may significantly underestimate the shallow dose. 

4.5	 REVIEW OF TBD-5 (ORAUT-TKBS-0015-5), PORTSMOUTH GASEOUS 

DIFFUSION PLANT – OCCUPATIONAL INTERNAL DOSE 

Relevant Background Information Pertaining to Recycled Uranium 

In 1952, the Atomic Energy Commission (AEC) approved the enrichment processing of 

production reactor returns or RU through the gaseous diffusion process. Between 1955 and 

1974, a total of 1,094.6 MTU of RU was processed as feed material that had been made at PGDP 

and the Oak Ridge Gaseous Diffusion Plant. Subsequently, smaller quantities of RU from other 

sources were also processed at PORTS, including 1.9 MTU of RU-UF6 extracted and produced 

by PORTS at the Decontamination and Uranium Recovery Building (Building X-705), Oxide 

Conversion Facility (Building X-705E), and Feed Production Facility (Building X-344). 

At the time, it was known that RU contains low concentrations of transuranics (e.g., Np-237, 

Pu-238/-239/-240, and Am-241) and select long-lived fission products, most notable Tc-99. 

When feed materials containing RU contaminants are processed, their contaminants tend to 

concentrate in either finished products or in liquid raffinates and waste streams, such as tower 

and filter ashes. For example, when TRU isotopes are fluorinated and fed into the cascades, they 

tend to follow the heavier isotope of U-238, which is differentially concentrated in the bottom of 

cascade cells near the feed points. Conversely, the lighter Tc-99 contaminant tended to 

concentrate in top cascade cells and cascade vent alumina traps (and after 1975 in MgF2 traps). 

Neither the presence of RU contaminants and their tendency to concentrate at select process 

locations nor their potential for worker exposure was fully recognized for many years at PORTS. 

In their investigative report (DOE 2000), DOE’s Office of Oversight provided the following 

observations, findings, and opinions: 

In 1957, radiological surveys at the Paducah Plant identified that neptunium-237 

was present in the enrichment cascade. Although the AEC recognized the 

potential for transuranic contamination of the cascades, it was not until a 1965 

appraisal that OR identified a potential problem with transuranics and fission 

products in X-705E and recommended studies to determine where these materials 

could concentrate in the process. Records reflect that PORTS then reviewed the 

potential problems posed by feeding reactor returns to the oxide conversion 

plant; however, detailed studies were not performed. PORTS correspondence 

also indicates that health physics staff did not fully understand the presence of 

transuranics and technetium-99, and appropriate analytical procedures were 

not developed as late as 1976. During the 1970s, PORTS health physics and 

Plant managers participated in preplanning for receipt and subsequent 

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processing of recycled uranium known to contain trace quantities of neptunium237, plutonium-239/240, and technetium-99. 

During the 1960s, the PORTS health physics group became concerned with 

increasing alpha radiation levels in process and support facilities at the site. 

While no records were identified to demonstrate that this issue was satisfactorily 

resolved, the period coincides with the processing of recycled uranium at the 

Paducah Plant. [Emphasis added.]

 From page 21: 

Some workers had extremely high intakes of uranium detected by bioassay or invivo testing that put them on work restriction for months or years. For example, 

in 1965 ten employees sustained lung exposures greater than one-half the 

permissible level, and eight were reported to the AEC as overexposures in 

accordance with AEC regulations. In addition, a worker who had a massive 

intake of UF6 in 1973 was still excreting uranium six months later, and two 

workers in 1965 were exposed to uranium levels high enough that, as late as 

1973, in-vivo testing showed greater than 50 percent of the maximum allowable 

body burden for uranium. Finally, one worker, still living, was put on 

permanent restriction in 1981, and his in-vivo monitoring before his 1985 

retirement still showed high uranium readings in his lungs. [Emphasis added.] 

 From page 69: 

Starting in 1975, Plant records reveal that elevated technetium and transuranic 

contamination was unexpectedly discovered in liquid process effluents from the 

X-705. Before then, radiological effluent monitoring was only conducted for 

uranium and indicator parameters. The PORTS environmental monitoring 

program did not include these contaminants, which were known by Plant 

management to have been introduced into PORTS industrial facilities from the 

processing of reactor returns and from Paducah production feed material. Based 

on the information collected, it does not appear that personnel responsible for 

environmental monitoring were aware of the presence of these contaminants at 

PORTS... 

...The increase in technetium-99 discharges occurred shortly after the initiation 

of equipment changeout in the X-330/X-326 process buildings. A technetium 

treatment system was proposed in the late 1970s and installed in the early 1980s 

to reduce the levels being discharged into the environment. 

By 1976, transuranics had also been identified in raffinates generated by the 

recovery of uranium from contaminated equipment and materials processed in 

X-705. These raffinates were discharged to the X-701B pond. Subsequent 

monitoring detected transuranics at significant levels in sludges from this pond 

and in the effluents from the pond to the east drainage ditch. Transuranics in the 

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effluent originated primarily in reactor-return materials processed in the X-705 

Building. As an outcome of these findings, a committee was formed in December 

1976 to study Plant-wide aspects of the transuranic contamination. 

...In 1979, isotopic analysis of two cascade deposits revealed relative high 

concentration of neptunium-237 (i.e., 55 percent and 60 percent of the total 

alpha activity in the samples was due to Np-237, respectively). However, there 

was no indication of a change in the radiological control program to address this 

issue, even though data was available to indicate that some level of transuranic 

contamination was present in the cascade. Transuranic sampling for work 

planning and control was not actively conducted until the 1990s. 

In June 2000, Bechtel Jacobs Company (BJC), in response to a request by the DOE, issued a 

report entitled Recycled Uranium Mass Balance Project Portsmouth, Ohio Site Report (BJC 

2000). The report provided a detailed analysis regarding quantities and flow paths for uranium 

reprocessed from spent fuel from plutonium and tritium production reactors at other DOE sites. 

Among the study objectives was the identification of work locations, which had the highest 

potential for exposures to RU contaminants. 

Locations and processes where worker exposure to TRU and Tc-99 was most probable were 

thought to include the following: 

•	 Decontamination and Uranium Recovery Building (X-705) and the Oxide Conversion 

Building (X-705E) 

These facilities recovered uranium from decontamination solutions and incinerator ash in 

preparation for conversion to UF6 for feed into the cascade. Between 1959 and 1961, 

uranium oxides from spent reactor fuel was processed. The BJC Report acknowledged 

that “. . . A limited amount of information is available that describes the recycled 

constituents of the oxide processed in this facility . . . [and] How much TRU was present 

in each year of operation is not known, however, sample results do verify that TRU 

contamination was present.” [Emphasis added.] 

DOE’s assessment of this facility included the following statements (DOE 2000): 

...Probably as the most hazardous operations at PORTS...[where] Processing 

of transuranic-contaminated material was not adequately anticipated in the 

original or subsequent designs or operation. Samples obtained after 

shutdown showing the presence and level of transuranic contamination in the 

facility indicate that worker airborne exposures could have exceeded the 

acceptable standards, especially given the apparent lack of discipline in 

respirator use. 

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• Buildings X-333, X-330, and X-326 

While TRUs were differentially concentrated/deposited in lower cascade cells near feed 

points, Tc-99 would sequentially concentrate in top cascade cells, as well as in alumina 

traps or MgF2 traps. Exposure to TRUs and Tc-99 (as well as uranium isotopes and their 

daughters) would have resulted from routine equipment repair and maintenance. 

However, the most intense maintenance and modification to the PORTS cascades 

occurred between 1972 and 1983 in behalf of a program called the cascade improvement 

program and cascade upgrade program (CIP/CUP). 

CIP/CUP replaced or upgraded key cascade components (e.g., converters, compressors, 

transformers, motors, etc.) in order to improve the diffusion process efficiency and 

reliability. 

• UF6 Cylinder Cleaning and Storage (Building X-745) 

UF6 cylinders of several diameter dimensions containing HEU RU were used at PORTS. 

The repeated filling of UF6 cylinders without washing/removing the non-volatile heel has 

the effect of concentrating TRU and FP in the heel. The degree to which the constituents 

concentrate depends on the amounts added with each filling and the fraction removed 

during each feeding. It is assumed that a fraction of cylinders experienced several cycles 

before they were cleaned. It was further assumed that some cylinders were never cleaned 

and contained unknown concentrations of TRU. 

Finding 4.5-1: Incorrect Values for Tc-99 were Derived for Reactor Tails 

Table 5.1.2.6-3 in TBD-5 provides activity values for TRU and Tc-99 in reactor tails in dpm/g of 

U on an annual basis. These values were supposedly derived from data contained in Table 3.2-1 

and Table 5.1-1 of the BJC 2000 report. For convenience, Tables 3.2-1 and 5.1-1 of the BJC 

2000 report are reproduced herein as Exhibits 4.5-1A and 4.5-1B. 

SC&A reviewed these data and concludes that Tc-99 values in Table 5.1.2.6-3 of TBD-5 are 

understated by several orders of magnitude, as illustrated below in a sample calculation. 

 Sample Calculation for Year 1955: 

(1) Exhibit 4.5-1A (BJC 2000 Table 3.2.1) identifies that for 1955, PORTS received 

105,873 kg of RU (or 1.059 × 108

 g RU). 

(2) Exhibit 4.5-1B (BJC 2000 Table 5.1-1) estimates that in 1955, the 105,873 kg of RU 

contained the following: 8 g of Np-237 and 3.45 kg of Tc-99. 

(3) From the t1/2 value of 2.14 × 106 years for Np-237, the specific activity of 7.14 × 10-4 

Ci/g is derived; and the 8 g of Np-237 in 105,873 kg RU yields the following activity: 

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February 6, 2008 0 SCA-TR-TASK1-0020 34 of 89 

(8 ) ( . x10 −4 g 714 Ci / g) Np-237 activity in RU = 105 873 000 , , g RU 

= 5.394 × 10-11 Ci/g RU

 = 53.94 pCi/g RU 

Np-237 = 118 dpm/g RU 

SC&A’s derived value of 118 dpm/g RU for Np-237 is fully consistent with NIOSH’s 

value of 1.2 × 102

 or 120 dpm/g RU, as given in Table 5.1.2.6-3 of TBD-5. 

(4) For Tc-99, however, SC&A derives an activity value as described below: 

With a half-life of 2.12 × 105 years, the specific activity of Tc-99 is equal to 0.01725 

Ci/g, and the total activity of 3.45 kg of Tc-99. Thus, the activity of Tc-99 in RU yield 

the following: 

(3450 g)( 0 01725 . Ci / g) Tc-99 activity/g RU = 105 873 000 , , g RU

59 5 .. Ci

= 

105 873 000 , , g RU

562 x10 −7 . Ci 

= 

g RU 

= 1,236,400 dpm/g RU
 

Tc-99/g RU = 1.24 × 106 dpm/g RU
 

SC&A’s value of 1.24 × 106

 dpm/g RU for Tc-99 is 24,000 times higher than NIOSH’s 

value of 56.1 dpm/g RU cited in Table 5.1.2.6-3 of TBD-5. A review of Tc-99 values cited 

in Table 5.1.2.6-3 indicates that all values suffer an error of this magnitude. 

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factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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EXHIBIT 4.5-1A 

Table 3.2-1 PORTS Receipts Summary (RU Only) 

(Source: BJC 2000) 

Shipping Facility Uranium 

Form 

Net Weight (kgU) 

FY 

1955 

FY 

1956 

FY 

1957 

FY 

1958 

FY 

1966 

FY 

1967 

FY 

1968 

FY 

1969 

FY 

1972 

Allied Chemical UO3 

Babcock & 

Wilcox UF6 

Division of 

International 

Affairs 

UF6 151 

UNH 7 39 

Fernald U3O8 

France UF6 65 

Germany UNH 

K-25 

UF4 865 

UF6 296,504 

UO2 418 

UO3 3,319 

NUMEC UF6 330 

Paducah UF6 105,873 54,649 6,156 64,311 567,620 

United Kingdom UNH 

USAEC Office 

Safeguards & 

Materials 

Management 

UF6 2,833 

Y-25 U3O8 

Grand Total 105,873 351,154 10,758 64,311 7 39 151 570,453 395 

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EXHIBIT 4.5-1A (Continued) 

Shipping Facility Uranium 

Form 

Net Weight (kgU) 

FY 

1973 

FY 

1974 

FY 

1975 

FY 

1976 

FY 

1976.5 

FY 

1977 

FY 

1978 Grand Total 

Allied Chemical UO3 1,376 1,403 1,295 4,074 

Babcock & 

Wilcox UF6 153 153 

Division of 

International 

Affairs 

UF6 151 

UNH 46 

Fernald U3O8 7,798 7,798 

France UF6 202 324 128 273 112 152 235 1,586 

Germany UNH 6,860 6,860 

K-25 

UF4 865 

UF6 296,505 

UO2 418 

UO3 3,319 

NUMEC UF6 330 

Paducah UF6 798,609 

United Kingdom UNH 7 7 

USAEC Office 

Safeguards & 

Materials 

Management 

UF6 2,833 

Y-25 U3O8 104 104 

Grand Total 1,578 324 1,538 16,226 112 409 235 1,123,658 

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EXHIBIT 4.5-1B 

Table 5.1-1. Annual Inventory of RU Constituents 

(Source: BJC 2000) 

Fiscal Year Np (g) Pu (g) Tc-99 (kg) Fiscal Year Np (g) Pu (g) Tc-99 (kg) 

1955 8.00 0.0 3.45 1977 49.62 0.23 61.61 

1956 16.00 0.0 6.90 1978 49.73 0.23 58.88 

1957 24.00 0.0 10.35 1979 49.73 0.23 56.78 

1958 32.00 0.0 13.80 1980 49.73 0.23 55.31 

1959 32.00 0.0 16.36 1981 49.73 0.23 52.83 

1960 32.00 0.0 18.92 1982 49.66 0.21 50.46 

1961 32.00 0.0 21.48 1983 49.46 0.16 49.15 

1962 32.00 0.0 24.03 1984 49.90 0.13 48.38 

1963 32.00 0.0 26.59 1985 48.90 0.13 47.68 

1964 32.00 0.0 29.15 1986 48.90 0.13 47.18 

1965 32.00 0.0 31.71 1987 48.90 0.13 46.84 

1966 32.01 0.0 34.27 1988 48.90 0.13 46.55 

1967 32.02 0.0 36.83 1989 48.90 0.13 46.33 

1968 35.47 0.02 39.75 1990 48.90 0.13 46.01 

1969 75.61 0.04 43.94 1991 48.90 0.13 45.82 

1970 75.61 0.04 46.50 1992 48.90 0.13 45.51 

1971 75.61 0.04 49.06 1993 48.90 0.13 44.40 

1972 75.72 0.05 50.32 1994 48.90 0.13 44.23 

1973 76.01 0.06 56.59 1995 48.24 0.12 43.89 

1974 45.42 0.06 63.99 1996 47.58 0.12 43.54 

1975 45.69 0.07 65.26 1997 46.92 0.11 43.20 

1976 48.03 0.17 64.72 1998 46.26 0.11 35.80 

Mid-1999 44.30 0.11 35.11 

Finding 4.5-2: Variable Bioassay Protocols were Employed that Significantly Affect the 

Interpretation of Urine Bioassay Data used for Dose Reconstruction 

At PORTS, worker exposure to uranium isotopes was assessed by measuring the concentration 

of uranium in urine samples. Important to dose reconstruction are urine bioassay data that reflect 

routine, special, and termination samples. 

Included among the many variables that affect the interpretation of urinalysis data for dose 

reconstruction is the time at which the urine sample is collected relative to the worker’s most 

recent exposure. This critical time interval apparently changed over time, as indicated by the 

following: 

DOE 2000 (page 37): 

. . . During the 1950s and 1960s, urine samples were typically analyzed for 

uranium, and in most cases for alpha activity. Typically, the sample collection 

procedure involved the collection of Monday morning urine specimens (the 

morning following two or more days off the job). This was non-conservative, and 

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the collection date evolved to a “Friday” sample during the 1970s and 1980s.... 

[Emphasis added.] 

The use of “Monday morning” and “the end of the shift on the last day of the employer’s 

workweek” urine samples are documented in the GAT Standard Operating Procedure, Industrial 

Hygiene and Health Physics Department 212 (GAT 1963b) and in GAT-S-54 Urinalysis 

Parameters (GAT 1985b) (see Exhibits 4.5-2A and 4.5-2B). However, SC&A was unable to 

determine from available data/documents whether this transition for urine collection protocols 

occurred at a fixed moment in time or oscillated over a period of years, as suggested by DOE 

(2000). 

NIOSH is fully aware that the uncertainty regarding the urine collection protocol will have a 

significant impact on the interpretation of bioassay data by IREP. 

A review of TBD-5 for PORTS, however, not only fails to mention the change in protocol for 

obtaining urine samples over time, but incorrectly implies that “Monday morning” samples were 

consistently used in the past up to the present time, as given in the following: 

...Routine samples were submitted on Monday of every week and recorded on 

form A-551. A special sample was given 4 hours after exposure and one voiding 

for suspected inhalation incidents. Total uranium analysis required 2 ml and a 

total alpha analysis required a 100-ml volume. Until around 1995, spot samples 

were the norm at PORTS. [Emphasis added.] 

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EXHIBIT 4.5-2A 

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EXHIBIT 4.5-2B 

(GAT 1986a) 

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EXHIBIT 4.5-2B (Cont.) 

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Finding 4.5-3: Current Guidance for Estimating Internal Exposure to RU Contaminants 

Unachievable and/or Inappropriate 

From the foregoing discussion, it is reasonable to conclude that any assessment of internal 

exposure to TRU and Tc-99 on the basis of in vitro urinalysis and in-vivo chest counting is 

severely limited. In behalf of these limitations, NIOSH provides the following guidance 

(Section 5.1.2.6, p. 13 of TBD-5): 

If specific source term information to which the employee has been exposed is 

available, the dose reconstructor should utilize that information. However, if no 

source term information is available, the values and parameters in 

Table 5.1.2.6-4 provide an adequate input to the process. Note that the activity 

fraction and contaminant ratio to uranium columns are based upon 1993 through 

1999 PORTS air sampling data and is therefore a snapshot of the temporal 

radionuclide matrix of the facilities. [Emphasis added.] 

SC&A regards NIOSH’s guidance for internal dose assessment involving TRU and Tc-99 (as 

well as for thorium isotopes and Pa-234m listed in Table 5.1.2.6-4) as either unachievable or 

inappropriate, as explained below. 

The term unachievable is a reference to NIOSH’s guidance to make use of “source term” 

information. Source term information implies the availability of data over time and location. 

Source term materials (e.g., filter ash, tower ash, cascade deposits, cylinder heels, pond sludges, 

MgF2 traps, raffinates, etc.) that may have concentrated select radionuclides and given rise to 

airborne contamination. Such data do not exist and will, therefore, not be available to the dose 

reconstructor. The term inappropriate is a reference to the use of general air sample data 

collected in the 1990s that only define activity fractions for select isotopes relative to uranium 

(see Table 5.1.2.6-4 of TBD-5). For use in dose reconstruction, these relative activity values 

defined in the 1990s must be applied to uranium bioassay data that may have been taken decades 

earlier and reflect processes/facilities/radiological conditions that have little to no relevance. The 

following examples illustrate SC&A’s concerns regarding the use of Table 5.1.2.6-4. 

 Example #1: DOE (2000), page 35, provides the following statement: 

... In 1979, isotopic analysis of two cascade deposits revealed relative high 

concentrations of neptunium-237 (i.e., 55 percent and 60 percent of the total 

alpha activity in the samples was due to Np-237, respectively). However, there 

was no indication of a change in the radiological control program to address this 

issue, even though data was available to indicate that some level of transuranic 

contamination was present in the cascade. Transuranic sampling for work 

planning and control was not actively conducted until the 1990s. [Emphasis 

added.] 

Such high concentrations of TRU contaminants are likely to have exposed cascade workers 

(Buildings X-333, 330, and 326) as well as maintenance/decontamination workers affiliated with 

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CIP/CUP (Buildings X-700, X-705, X-705E, and X-720). Inspection of Table 5.1.2.6-4 in TBD5 identifies Np-237 activity fractions that are several orders of magnitude lower. 

 Example #2: GAT-521-75-113 (GAT 1975) and BCJ (2000): 

In the interdepartmental correspondence enclosed herein as Exhibit 4.5-3A, cascade 

samples were analyzed for Tc-99 in the inlet lines and valves at a concentration of 20% 

by weight. 

Even higher concentrations were cited in the BJC 2000 Report, which stated the 

following (pp. 63–64): 

The Tc-99 contamination in cascade equipment has been an acknowledged 

problem since the mid-1970s. Some materials removed from the cascade at 

areas known to concentrate Tc-99 have been as high as 40% by weight Tc-99. 

At these concentrations, the relative activity of Tc-99 to uranium is orders of magnitude 

higher than values cited in Table 5.1.2.6-4. 

In summary, credible bioassay data for TRU and FP contaminants contained in recycled 

uranium are not available, and the default methodology imbedded in Table 5.1.2.6-4 is 

inappropriate and not claimant favorable. 

 Example #3: From document No. 74 – Summary of Materials on Plantsite from TRU Activities 

at X-705 (see Exhibit 4.5-3B) 

Exhibit 4.5-3B identifies the analysis of ash in “container F861667.” This ash material 

was assayed at 2.9% U-235 enrichment. As such, the specific activity of uranium in the 

ash is estimated at 1.53 μCi/g: 

• Sp. Act of Uenrich = (0.4 + 0.38E + 0.0034E2

) μCi/g 

= (0.4 + 1.10 + 0.0286) μCi/g 

= 1.53 μCi/g U 

• Thus, the PuTotal of 1,970,000 dpm/g U is equal to 0.895 μCi/g U and represents a 

fractional activity 5.85 × 10-1 relative to uranium. 

The above-derived value of 5.85 × 10-1 is more than 2,000 times the generic value of 

2.64 × 10-4 recommended for the X-705 and X-705E location defined in Table 5.1.2.6-4. 

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EXHIBIT 4.5-3A 

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EXHIBIT 4.5-3B 

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Finding 4.5-4: Generic Default Value for Uranium Enrichment is Inappropriate/Claimant 

Unfavorable 

Worker exposure to uranium at PORTS was primarily assessed by means of urinalysis. Because 

exposures at PORTS may have involved depleted, natural, low-enriched, enriched, highly 

enriched, and very highly enriched uranium, an assessment of both chemical toxicity and 

radiological exposure required that two independent bioassays be performed concurrently: 

(1) the concentrations of elemental uranium in urine was determined by a standard fluorimetric 

procedure; and (2) the specific activity of the uranium isotopic mixture was determined by gross 

alpha activity. An essential step in the gross alpha urinalysis is the chemical isolation/extraction 

of uranium from a urine sample. 

For dose reconstruction purposes, it is the second or gross alpha urinalysis that is critical. In a 

1966 GAT Report entitled, Determination of Uranium Alpha Activity in Urine (GAT 1966), the 

authors described a new and efficient extraction method that would replace the previous/existing 

methods, as given in the following: 

... Previously available procedures for determining alpha activity, however, have 

not been entirely satisfactory; the main problem has been the long, tedious 

chemical operations used to separate a sufficient quantity of uranium from urine 

for satisfactory alpha counting. This situation is exemplified in the previously 

used method of Whitson and Kwasnoski in which uranium is separated from 

100 ml of urine by (1) destruction of organic matter by wet oxidation and ignition, 

(2) two precipitations with oxalate, and (3) electroplating on a nickel disk. 

Regardless of the time period and which method was used for gross alpha analysis of urine 

samples, not all urine fluorimetric bioassays were concurrently evaluated for gross alpha activity. 

For those instances in which fluorimetric bioassay data are not accompanied by gross alpha data 

Section 6.7 on page 24 of TBD-5 informs the dose reconstruction to assume “. . . uranium 

enrichment – 3.5% unless HEU of 93% is suspected.” 

It should be noted that a 3.5% enrichment has the specific activity of 2.2 pCi/μg of uranium (or 

2.2 μCi/g of uranium). 

SC&A questions the default value of 3.5% enrichment with its attendant specific activity of 

2.2 μCi/g of U as generically applicable without consideration of time and location/job function. 

In the 1986 GAT-S-60 Report (GAT 1986b), data are presented that identify huge fluctuations in 

the specific activity values over time and by job categories. Provided herein as Exhibits 4.5-4A 

and 4.5-4B are summary bioassay data for all monitored workers for years 1965 through 1985. 

While page 1 of Exhibit 4.5-4A identifies gross alpha data (in dpm/100 ml urine), Exhibit 4.5-4B 

identifies uranium concentrations (in mg/liter of urine). By comparing these data, an estimate of 

the specific activity can be readily determined, as illustrated below: 

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 Example Calculations of Average Specific Activity for All Monitored Workers

 For 1965: 

Average α activity for routine urine = 5.67 dpm/100 ml

 = 56.7 dpm/liter 

= 25.7 pCi/liter 

Average U conc. for routine urine = 1.87 × 10-3 mg/liter 

= 1.87 μg/liter 

25 7 . pCi / liter 

Specific Activity = 187 . / μg liter

 = 13.8 pCi/μg 

For 1969: 

Average activity for routine urine = 2.46 dpm/100 ml

 = 24.6 dpm/liter 

= 11.2 pCi/liter 

Average U conc. for routine urine = 9.65 × 10-5 mg/liter 

= 9.65 × 10-2 

μg/liter
 

11 2 .. pCi / liter 

Specific Activity = − 9 65 x10 2 . μg / liter 

= 116 pCi/μg U 

The empirically derived yearly average values of 13.8 pCi/μg U and 116 pCi/μg U for 1965 and 

1969, respectively, are more than 6 and more than 50 times higher than NIOSH’s recommended 

generic default value of 2.2 pCi/μg U. 

When bioassay data are further segregated by job function, the disparity for select job functions 

is further amplified. Exhibits 4.5-4C and 4.5-4D identify bioassay data for workers classified as 

Material Handling and Sampling. For 1965, the average annual specific activity of excreted 

uranium is calculated at 120 pCi/μg U. Even higher values are derived for Chemical Operators, 

whose data are given in Exhibits 4.5-4E and 4.5-4F. For 1969, the annual average specific 

activity for routine urine samples is estimated at 147 pCi/μg U, which is nearly 67 times the 

default value of 2.2 pCi/μg U. 

Derived specific activity levels of this magnitude suggest that worker exposure involved high to 

very highly enriched uranium. Thus, the generic default assumption of 3.5% enrichment must 

clearly be regarded as inappropriate and claimant unfavorable. 

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EXHIBIT 4.5-4A 

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EXHIBIT 4.5-4B 

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EXHIBIT 4.5-4C 

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EXHIBIT 4.5-4D 

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EXHIBIT 4.5-4E 

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EXHIBIT 4.5-4F 

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Finding 4.5-5: Inappropriate Use of MDC Values for Deriving Missed Dose of Monitored 

Workers 

Section 5.7.2 of TBD-5 contains the following guidance for dose reconstruction of “Monitored 

Workers with Nothing Detected in Bioassay:” 

For monitored workers with no positive results, a triangular distribution is used, 

with the mode determined using half of the MDC value and the maximum using 

the MDC as input into the dosimetry codes. 

Table 5.1.1-1 of TBD-5 identifies that for 1954–1995, the MDC (as well as the “Minimum 

Recorded Level”) for gross alpha counting was 10 dpm/liter (or 1 dpm/100 ml). In Section 

5.1.2.1, (In Vitro Bioassay for Uranium) of TBD-5, however, the following statements are cited 

in behalf of gross alpha counting of urine samples: 

... the alpha activity was measured on a proportional counter in α

counts/min/100 ml urine {1 dpm/100 ml minimum detectable concentration 

(MDC)}. 100 ml of urine was needed for this technique. A result of 5 dpm/100 ml 

was considered the reporting level for the time-period of 1954–1993 (GAT, 1955; 

GAT, 1985a; Hill and Strom, 1993). Review of claim records reveals that 

sometimes values less than 5 dpm were recorded, probably down to the MDC of 

1 dpm/100 ml. 

SC&A reviewed the above-cited reference (GAT 1985a), which verified the fact that for gross 

alpha counting of urine sampling, the minimum recorded concentration/level at 5 dpm/100 ml 

(or 50 dpm/liter) is 5 times higher than the MDC level of 1 dpm/100 ml (or 10 dpm/liter) (see 

Exhibit 4.5-5A). 

SC&A concludes that TBD-5 contains contradictory/erroneous data, and guidance that instructs 

the use of MDC values instead of minimum recorded values is inappropriate and claimant 

unfavorable. 

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EXHIBIT 4.5-5A 

(Source: GAT 1985a) 

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Finding 4.5-6: MIVRML Chest Counts for the Detection of Insoluble Isotopes of Uranium, 

TRUs, and Fission Products are Subject to Significant Limitations and Uncertainties 

Relevant Background Information 

Fourteen years after startup, an in-vivo body counting program was initiated in 1965 to monitor 

for insoluble enriched uranium. Initially, until November 1967, all counting was done at the 

in-vivo facility of the Y-12 plant in Oak Ridge. For 1965 and 1966, only 27 and 30 chest counts, 

respectively, were performed on PORTS workers. Since November 1967, chest counts were 

performed by the Mobile In-Vivo Radiation Monitoring Laboratory (MIVRML). Although some 

PORTS workers were provided chest counts at the MIVRML while the mobile unit was at the 

National Lead of Ohio or FMPC facility, most chest counts of PORTS workers were done at the 

plant during the semi-annual visits. 

The MIVRML employed two very large sodium iodide crystal detectors measuring 9 inches in 

diameter by 4 inches thick. While such large/thick crystals offer maximum counting efficiency 

that is highly desirable for whole-body counting of high-energy photons of fission/activation 

products (e.g., Cs-137, Co-60), their use in detecting low-energy photons is severely 

compromised by their 4-inch thickness. This limitation in sensitivity is due to the fact that for 

large/thick crystals, the low-energy photons fall into that region of Compton-scattered photons 

that is maximal. 

For the MIVRML, detection of U-235 was based on the 186 keV photon, while detection of 

U-238 relied on the 63 keV and 93 keV photons emitted by the Th-234 daughter product. 

Important to note is that these two photons also have low yields only 3.5% and 4%, respectively. 

The detection of Np-237 was based on multiple photon emissions (Np-237 and its radioactive 

daughter Pa-233); and detection of Tc-99 (a pure beta emitter) was based on the production of 

Brehmstrahlung. This system was used until 1991, when it was replaced by the Helgesson 

Phoswich detector system. With its 0.5-inch thick crystals, detection limits significantly 

improved. However, by this time, the potential for exposure to insoluble uranium compounds 

that largely involved Buildings X-705 and Building X-705E had been reduced/ eliminated. 

In May 2000, DOE’s Office of Oversight issued a report entitled Independent Investigation of 

the Portsmouth Gaseous Diffusion Plant. In behalf of in-vivo chest counting, the report stated 

the following: 

Page 4: 

... In 1965, an in-vivo body counting program was initiated to monitor for 

insoluble enriched uranium, a material for which the urinalysis program was not 

sufficiently sensitive or reliable. Studies performed in 1990 indicated that the invivo counter’s capability for analyzing transuranics was questionable, making it 

difficult to demonstrate that all internal exposures have been accurately detected 

and assessed. 

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Pages 38–39: 

In a report prepared for Martin Marietta Utility Services in 1990, the 

effectiveness of the mobile whole body counter was evaluated for analysis of 

uranium, neptunium, plutonium, and americium. Additionally, a review was 

conducted of historical lung counting data from Martin Marietta Utility Services 

sites, with particular emphasis on neptunium-237. A summary of the findings 

indicated that the counter’s capability for analysis of those radionuclides, with 

the exception of uranium-235, was somewhat questionable due to system 

hardware limitations (i.e., use of sodium iodide detectors, resolution of spectra 

insufficient to identify peaks in the presence of background radiation, efficiency 

calibrations did not use multiple source strength measurements for isotopes 

other than uranium-235). The studies of historical data indicated difficulties, 

including the inability to retrieve the appropriate data, lack of system access, and 

insufficient documentation. The root cause for most of the problems could be 

attributed to physical limitations of the system, lack of understanding of these 

limitations, and the lack of adequate training. Incomplete isotopic and uranium 

solubility characterization, coupled with design and analytical limitations, has 

impacted the Plant’s ability to demonstrate that all internal exposures have been 

accurately detected and assessed. [Emphasis added.] 

SC&A not only agrees with these conclusions that place root cause on the physical limitations of 

the MIVRML, but has previously voiced these same concerns in behalf of other DOE facilities 

that provided chest counting by means of the MIVRML (see draft report, Review of the Feed 

Materials Production Center (FMPC) Special Exposure Cohort (SEC) Petition-00046 and the 

NIOSH SEC Petition Evaluation Report (SCA 2007). 

Furthermore, the system’s failure to accurately assess U-238 renders the only “reliable” 

measurements of U-235 useless for the following reasons: 

(1) In the absence of an accurate U-238 measurement, the activity of U-235 provides no 

information regarding the enrichment of uranium and, therefore, the variable activity and 

dose contributions of U-234 and U-238. 

(2) For even modestly enriched uranium, it is U-234 that dominates activity (and therefore 

dose), as shown for the following levels of enrichment: 

U-235 % Activity 

Enrichment U-234 U-235 U-238 

Depleted U 15.5 1.1 83.4 

U Natural 48.9 2.3 48.9 

2.0% 64.8 4.1 31.1 

3.5% 81.8 3.4 14.7 

93% 96.8 3 0.03 

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Exhibit 4.5-6A provides annual summary in vivo data in behalf of all monitored workers for the 

years 1965 through 1985. A visual inspection that compares columns #3 and #4 clearly 

demonstrates the questionable values of “total U results” relative to U-235 measurements (GAT 

1986a). 

In conclusion, SC&A regards in-vivo bioassay data of limited value. 

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EXHIBIT 4.5-6A 

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4.6	 REVIEW OF TBD-6 (ORAUT-TKBS-0015-6) PORTSMOUTH GASEOUS 

DIFFUSION PLANT – OCCUPATIONAL EXTERNAL DOSE 

Relevant Summary Data 

Between 1954 and 1980, exposure to external photon and beta radiation at PORTS was 

performed by means of a 2-element film dosimeter—an open window and a shielded element. 

This film badge dosimeter contained Kodak Type-2 film with two emulsions for normal 

conditions. The more sensitive emulsion responded to doses between 30 mR and 2,000 mR, 

while the less sensitive emulsion had a range of 5 R to 300 R and was reserved for 

accidental/criticality exposures. Important to note was that the film dosimeters were processed 

in-house. 

Starting in 1981, PORTS replaced the film dosimeters with a series of TLDs that included a 4­

element Harshaw TLD, a 4-element Panasonic TLD, and a 4-element International Chemical and 

Nuclear Corporation (ICN) TLD. 

Personnel monitoring for neutron exposures at PORTS did not begin until 1992, and routine 

personnel monitoring for neutrons did not occur until 1997. Personnel neutron dosimeters 

employed Li-6 phosphor in a design that measured albedo neutrons. 

Specific dates for use of individual dosimeters, their exchange frequency, limits of detection, and 

selective use are summarized in Table 6-3 of TBD-6. For convenience, Table 6-3 is reproduced 

herein as Exhibit 4.6-1A. 

A substantial portion of TBD-6 is committed to supportive background information regarding 

dosimeter performance specifications/uncertainties, gaps in monitoring data, and deficiencies in 

monitoring practices. Data gaps and monitoring deficiencies acknowledged in the TBD include 

the following: 

(1) Prior to 1960, select persons/groups with the highest exposure potential were monitored; 

this implies that other persons (albeit with lower exposure potential) were not monitored. 

(2) The presence of substantial quantities of beta emitters (e.g., Tc-99, Th-234, and Pa-234) 

have the potential for high external shallow doses/skin doses, as well as from skin 

contamination; however, extremity monitoring was not performed. 

(3) Neutron exposures may have occurred throughout the years of PORTS operations, but 

routine monitoring was not provided until the 1990s. 

In summary, these data gaps and monitoring deficiencies fall into the categories of “missed and 

unmonitored” dose and were addressed in TBD-6 in the following subsections: 

• Section 6.5.1: Missed and Unmonitored Shallow Dose 

• Section 6.5.2: Missed and Unmonitored Photon Dose 

• Section 6.5.3: Missed and Unmonitored Neutron Dose 

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EXHIBIT 4.6-1A 

Table 6-3. PORTS Dosimeter Type, Period of Use, Exchange Frequency, LOD, and Potential Annual 

Dosimeter Period of Use Exchange Frequency Laborato

Beta/photon dosimeters PORTS film 2-elementb 

9/22/54–7/16/57 Weekly (n=52) {selected groups} 0.0

7/17/57–9/30/58 Biweekly (n=26) {selected groups} 0.010/01/58–4/8/59 Weekly (n=52) {chemical operators and material handlers} 

Monthly (n=12) {remainder of selected groups} 

0.0

0.0

4/9/59–7/31/60 Every 4 wk (n=13) {all selected groups} 0.0

8/1/60–7/5/64 Monthly (n=12) {all selected groups} 

Quarterly (n=4) {all other employees} 

0.0

0.0

7/6/64–12/28/69 Quarterly (n=4) {all employees} 0.012/29/69–12/30/73 Quarterly (n=4) {selected employees} 0.012/31/73–6/29/75 Quarterly (n=4) {selected employees} 

Semiannual (n=2) {unselected employees} 

0.0

0.0

6/30/75–12/31/80 Quarterly (n=4) {selected employees} 

Monthly (n=12) {selected female employees only} 

0.0

0.0

PORTS Harshaw 2276 4-element TLD without window 1/1/81–12/31/82 Monthly (n=12) {all monitored} Quarterly (n=4) {all monitored} 0.015d 0

PORTS Harshaw 2276, 8000, 8800 4-element TLD with window 

1/1/83–12/31/98 {1/1/93–12/31/96 for BJC employees} Quarterly (n=4) 0.010e(0.0

ICN TLD 760 1/1/99–present {USEC employees} Quarterly (n=4) 0.01f(0.03 

ORNL Panasonic 8805/8806 4-element TLD with window 1/1/99–present {BJC employees} Quarterly (n=4) 0.01h(.03 Neutron dosimeters PORTS TLD albedo dosimeter {USEC and BJC} 

1/1/1992–12/31/94 {unmoderated Cf-252 calibrated} Quarterly (n=4) 0.0

1/1/95–12/31/96 {moderated Cf-252 calibrated} Quarterly (n=4) 0.0

ICN TLD 760 {USEC} 1/1/97–present {moderated Cf-252 calibrated} Quarterly (n=4) 0.0Y-12 Panasonic TLND {BJC employees} 1/1/97–12/31/98 Quarterly (n=4) 0.0ORNL Panasonic TLND 8806 4-element TLD {BJC employees} 1/1/1999–present Quarterly (n=4) 0.0

a. Maximum annual missed dose (NIOSH 2002): For photon/beta missed dose = LOD/2 × n (frequency (p. 18); for neutron missed dose = LOD/2 × n(frequency) (p. 29). b. Kodak personnel Type 2 film with gold sandwiched with cadmium for high-energy gamma, OW with aluminum for low-energy gamma and beta. LOD for SDE and DDE are the same (0.03 . GAT (1963a), 30 mrem reporting level for gamma and beta. d. (Wagner 2003). e. Bassett (1986a, p. 3). f. ICN (2003). g. Souleyrette (2003). h. McMahon (2003). i. Cardarelli (1997). 

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Missed and Unmonitored Shallow Dose 

For missed shallow dose (SDE), Section 6.5.1 of the TBD recommends the use of n(LOD/2) and 

resultant values shown in Table 6-29. For the 2-element film dosimeter, the LOD for shallow 

dose of 30 mrem is identified. 

For unmonitored shallow dose, the TBD states that “. . . the dose reconstructor should consult 

ORAUT-OTIB-0040 for instructions.” The TBD further states that the empirical PORTS SDE 

data provided in Table 6-31 of the TBD “. . . are for informational purposes only.” 

Finding 4.6-1: The Assumed LOD Value for SDE Associated with the Two-Element Film 

Dosimeter used Between 1954 and 1980 Lacks Technical Support and is Not Claimant 

Favorable 

The following information is provided in Section 6.3.2.1.1 of TBD-6: 

The film dosimetry program began in 1954. The dosimeter description from the 

documentation obtained is cryptic. A Description of Co-Operative Work 

Assignments in Industrial Hygiene and Health Physics (Wooldridge 1964) 

describes the dosimeter as a “film badge with Kodak Type-2 personal monitoring 

film combined with aluminum, cadmium, and gold filters for beta-gamma, low 

energy gamma, and high energy gamma radiation. There are also sulfur and gold 

filters for neutron exposures.” This document indicates a detection range from 30 

to 2,000 mrem... [Emphasis added.] 

Exhibit 4.6-1B identifies the source of information used by NIOSH to base their decision to 

employ the LOD value of 30 mrem for the shallow dose. 

SC&A interprets the statement in Exhibit 4.6-1B “. . . The present film will detect from 

approximately 30 millirem...” to mean that 30 mrem was the minimum reportable dose (as 

opposed to LOD). Moreover, the detection level of 30 milliroentgen is likely to refer to the deep 

dose equivalent (DDE). In fact, NIOSH shares this interpretation as indicated in footnote b of 

Table 6.3 shown in Exhibit 4.6-1A, where NIOSH states the following: 

Kodak personnel type 2 film with gold sandwiched with cadmium for high-energy 

gamma, OW with aluminum for low-energy gamma and beta. LOD for SDE and 

DDE are the same (0.03 rem), as the reporting level. [Emphasis added.] 

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EXHIBIT 4.6-1B 

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SC&A disagrees with NIOSH’s unsupported assumptions regarding the LOD value of 30 mrem 

for the following reasons: 

(1) No documentation or empirical data exist that defines the SDE LOD value for the 

PORTS 2-element film dosimeter. 

(2) The minimum reporting dose (MRD) is not equivalent to the LOD. 

(3) Footnote b of Table 6-3 of the PORTS TBD-6 (see Exhibit 4.6-1A) makes reference to 

“...OW with aluminum for low-energy gamma and beta,” which suggests that the open 

window (OW) of the 2-element dosimeter may have used an aluminum filter, which 

would limit its sensitivity to the low-energy (i.e., 294 keV Tc-99) betas. 

(4) Barring the presence of an aluminum filter, the standard Kodak Type-2 film dosimeter 

used at other DOE facilities is consistently characterized as having an SDE LOD value 

of 40 mrem. 

(5) The ability to assess shallow dose from beta radiation may have further been 

compromised by a practice of wearing the badge “inside the anti-Cs,” as suggested in 

handwritten notes by a cascade process operator in 1986 that coincided with the 

CIP/CUP program. Enclosed as Exhibit 4.6-1C are pages 1, 2, 5, and 6 of the 

handwritten notes. (Note: Select passages are highlighted by arrows on the left-hand 

margin.) 

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EXHIBIT 4.6-1C (Handwritten Notes - Page 1) 

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EXHIBIT 4.6-1C (Handwritten Notes - Page 2) 

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EXHIBIT 4.6-1C (Handwritten Notes - Page 5) 

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EXHIBIT 4.6-1C (Handwritten Notes - Page 6) 

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Finding 4.6-2: Unmonitored SDE Estimates Derived from Co-Worker Data Suffer 

Deficiencies that are likely the Result of Dosimeter Design Limitations and/or Dosimeter 

Process Policies 

ORAUT-OTIB-0040 (External Coworker Dosimetry Data for the Portsmouth Gaseous Diffusion 

Plant) is to be used for assignment of unmonitored dose. Key to the assignment of unmonitored 

dose is the use of Portsmouth monitored coworker data for which 50th and 95th percentile dose 

estimates have been derived on a annual basis. In addition to recorded dosimeter doses, 

coworker doses were modified to account for missed doses (assumedly all individual doses 

recorded as zero were replaced by LOD/w or 30 mrem/2 or 15 mrem for both SDE and DDE. 

Table 8-2 of OTIB-0040 provides the adjusted Portsmouth coworker dose data and is herein 

reproduced as Exhibit 4.6-2A. 

Inspection of Exhibit 4.6-2A shows that with few exceptions, the non-pentrating (or beta) 

component that would be expected to have registered on the OW portion of the film dosimeter 

was on average essentially zero (or nonexistent). Due to the presence of substantial amounts of 

Tc-99, as well as several radioactive daughter products of uranium that moreover concentrated, 

the near absence of measurable beta radiation up until 1980 is unexpected and must, therefore, be 

viewed with some suspicion. Possible explanations may include issues raised in Finding 4.6-1. 

Another issue of concern that affects the ability to account for both missed and unmonitored 

exposures involves a statement contained in the 2000 report issued by the Bechtel Jacobs 

Company LLC (BJC 2000). On page 68 of the report entitled Recycled Uranium Mass Balance 

Project Portsmouth, Ohio Site Report, the following statement appears: 

Worker monitoring began in 1954 with the Film Badge and Bioassay Programs. 

Workers with the potential for external radiation were provided film badges for 

monitoring. However, not all workers were provided film badges, and not all 

badges issued to workers were read. This changed in the mid-70s when the film 

badges were replaced with TLD badges. . . . Some badges are not read unless 

there is cause to believe a significant dose may have been recorded. [Emphasis 

added.] 

Page 68 of the BJC (2000) report is enclosed as Exhibit 4.6-2B. It should be noted that this 

exhibit also provides a relevant discussion about Tc-99. 

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EXHIBIT 4.6-2A 

Year 

1954 

1955 

1956 

1957 

1958 

1959 

1960 

1961 

1962 

1963 

1964 

1965 

1966 

1967 

1968 

1969 

1970 

1971 

1972 

1973 

1974 

1975 

1976 

1977 

1978 

1979 

1980 

1981 

1982 

1983 

1984 

1985 

1986 

1987 

1988 

1989 

1990 

1991 

1992 

Gamma 

95th% 

1.736 

1.104 

0.945 

0.714 

0.857 

1.164 

0.283 

0.240 

0.332 

0.360 

0.262 

0.140 

0.136 

0.122 

0.338 

0.281 

0.459 

0.281 

0.367 

0.407 

0.337 

0.474 

0.415 

0.365 

0.414 

0.181 

0.307 

0.120 

0.112 

0.053 

0.053 

0.045 

0.058 

0.063 

0.057 

0.063 

0.075 

0.057 

0.054 

Gamma 

50th% 

0.780 

0.874 

0.799 

0.615 

0.574 

0.591 

0.195 

0.180 

0.180 

0.180 

0.120 

0.060 

0.060 

0.060 

0.060 

0.085 

0.158 

0.060 

0.078 

0.077 

0.060 

0.078 

0.060 

0.078 

0.077 

0.060 

0.060 

0.090 

0.090 

0.020 

0.020 

0.020 

0.020 

0.020 

0.020 

0.020 

0.022 

0.020 

0.020 

Non-pen 

95th% 

0.055 

0.048 

0.016 

0.335 

0.434 

0.579 

0.123 

0.086 

0.094 

0.051 

0.021 

0.050 

0.010 

0.005 

0.007 

0.011 

0.047 

0.065 

0.173 

0.087 

0.033 

0.114 

0.050 

0.050 

0.223 

0.174 

0.001 

N/A1 

N/A1 

0.112 

0.080 

0.083 

0.082 

0.083 

0.087 

0.091 

0.098 

0.101 

0.109 

Non-pen 

50th% 

0.000 

0.000 

0.000 

0.000 

0.031 

0.040 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.005 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.000 

0.016 

0.000 

0.000 

N/A1 

N/A1 

0.060 

0.060 

0.060 

0.053 

0.046 

0.060 

0.047 

0.055 

0.050 

0.060 

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EXHIBIT 4.6-2B 

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Finding 4.6-3: External Exposures to Localized Skin and to Extremities were Inadequately 

Monitored at PORTS, and Guidance to Dose Reconstructors is too Subjective and 

Arbitrary 

Numerous activities at PORTS subjected workers to external radiation fields in which 

extremities (i.e., hands and feet) would have received exposures that were substantially higher 

than shallow doses recorded by a personnel dosimeter worn on the anterior portion of the chest. 

Activities with potential for high extremity doses would have included those involving oxide 

conversion, ash handling, routine maintenance and CIP/CUP efforts, and equipment 

decontamination. A substantial component of extremity doses would have been contributed by 

beta radiation, as noted in DOE 2000 (page 36): 

Routine whole body beta exposures in excess of PORTS investigative levels 

existed primarily in areas where uranium daughter products tended to 

concentrate. ...Exposure evaluations during the mid to late 1950s indicated 

numerous instances of workers being placed on work restrictions based on whole 

body exposures that were determined to be in excess of PALs [Plant Allowable 

Limits]. Documents also indicated that before the mid-1980s, Goodyear Atomic 

Corporation had never performed extremity monitoring for any operation or 

work activity. Documents indicated that various valves associated with pigtail 

operations had recorded beta readings as high as 1 rad/hour. Feed production 

plant ash receiver areas had floor readings of 5 rad/hour beta. Operators 

routinely handled these valves and equipment in X-705 and other locations where 

significant hand exposures could occur. [Emphasis added] 

In addition to unmonitored extremity exposures from external sources, unrecorded skin and 

extremity exposures would also have resulted from personnel contamination involving bare skin, 

as well as from clothing. (It should be noted that select radionuclides, inclusive of Tc-99, are not 

easily removed from the skin by ordinary washing. Failure to detect localized contamination 

would, therefore, yield long residence half-times and with time, produce substantial localized 

skin doses.) 

To accommodate extremity monitoring deficiencies and potential exposures resulting from skin 

contamination, TBD-6 (pages 41–43) provides several hypothetical “examples,” including 

source term models. 

SC&A believes that for dose reconstruction of skin/extremities, the current guidance is too 

subjective and arbitrary. 

Finding 4.6-4: Before 1992, PORTS failed to monitor workers for neutron exposures. 

Current guidance to account for unmonitored neutron exposures is incomplete. 

Based on the large quantities of uranium compounds at PORTS, it is axiomatic that neutron 

exposures have existed throughout PORTS as a result of spontaneous fission and, more 

importantly, from the alpha/neutron reactions of UF6. Neutron exposure rates were likely 

highest in areas were uranium was stored (i.e., cylinder yards), handled (i.e., feed and withdrawal 

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areas), or deposited (i.e., within the cascade). Important variables affecting neutron production/ 

exposure include the level of uranium enrichment, geometry of the source term, and neutron 

moderation. 

Concern for neutron exposures associated with unplanned accumulation of fissile material 

sufficient to initiate criticality was recognized from the beginning. On October 5, 1954, 

Goodyear Atomic Corporation issued a report entitled, Significant Incident Report: 

Condensation of UF6 in the Cascade during Initial Start-Up Operations (GAT 1954). This 

report noted that on September 11, 1954, conditions were favorable to condensation of UF6 gas 

in surge drums that were used as a reservoir during startup operations. The report also noted, 

“...that one of the chief critical mass hazards in a gaseous diffusion plant is the unplanned 

accumulation of a mass of fissionable material sufficient to initiate a critical reaction.” 

In a 1957 Committee Report (FR-434-06-0035 Box-1), a total of 105 “false alarms” were 

recorded in a single year by 29 criticality instruments at PORTS. At PORTS, criticality 

instruments were Argon Gamma Graphs, which respond to transient spikes in gamma radiation 

and, therefore, only indirectly record potential criticality events. Furthermore, these instruments 

had five different alarm settings. An investigation of causes for the 105 alarms provided answers 

to 82 false alarms. However, for the remaining 23 criticality alarms in 1957, no plausible 

explanation could be identified. SC&A believes that near-criticality conditions cannot be 

ruled out as a cause. 

Sub-critical accumulations of UF6 were routinely observed at PORTS. For example, in a 

September 9, 1990, report entitled, Significant Non-Critical Incident No. 49: Deposit of 

Uranium Material in Cells X-27-3-9 and X-27-3-11 (GAT 1990), the following quantities of 

accumulated materials were discovered by means of portable Ludlum gamma spectrometer, as 

given in the following reported findings: 

•	 ... the deposits in X-27-3-9 in the upper part of the cell were reduced from 

22 pounds to 12 pounds UO2F2 in the compressor and from 38 pounds to less 

than 13 pounds UO2F2 in the coller. At an assay of 15 percent U-235 

(assumed to be the assay of the deposit) the ordinary safe mass is 10 pounds 

UO2F2 and the minimum critical mass is 27 pounds UO2F2. [Emphasis 

added.] 

•	 The deposit in the X-27-3-11 converter was reduced from 30 pounds UO2F2 

to less than 2 pounds UO2F2 .... 

Of relevance here is the fact that dosimetry programs at PORTS from 1954 to 1992 failed to 

monitor workers by means of calibrated personnel neutron dosimeters. Reference to the term 

“calibrated” requires additional explanation. When film dosimeters were replaced by a 4­

element TLD in 1981, three of the chips consisted of Li-7 (which responds to betas and photons), 

while the 4th chip of Li-6 responds to betas, photons, and neutrons. However, this 4-element 

TLD was not calibrated or evaluated for neutron exposures. Routine processing of these 

personnel dosimeters for beta/gamma exposures in some instances yielded highly abnormal chip 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

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ratios that either reflected a damaged dosimeter or neutron exposures as high as 4.8 rem (DOE 

2000). 

Unfortunately, these abnormal/inexplicable chip ratios were dismissed by PORTS as damaged 

chips. The combined failure to monitor workers for neutrons and dismissal of unexplained chip 

ratios resulted in a 1996 request for NIOSH to conduct a Health Hazard Evaluation (HHE) at 

PORTS. The principal objectives of the evaluation were two-fold; (1) assess historical neutron 

exposures at PORTS, and (2) resolve the anomalous chip ratio values. In 1997, NIOSH released 

its HHE (Cardarelli 1997). With regard to Objective #2, the HHE provided the following 

explanation: 

 From page 7 on the nature of slow cookers: 

A slow buildup of uranium material within the cascade causes a slight increase in 

the production of neutrons . . . This phenomenon of a slow build-up of uranium 

material that approaches criticality has been termed by the gaseous diffusion 

industry as a “slow cooker.” In essence, a slow cooker is a mass of uranium in 

which there is a multiplication of neutrons but at a rate below the critical 

threshold. Slow cookers are directly associated with uranium deposits which 

have routinely occurred at the PORTS since the plant’s inception. . . . [Emphasis 

added.] 

Neutron doses specifically attributable to this phenomenon could not be addressed in this 

evaluation for the following reasons: 

 From page 11 on the inability to retrieve past data. 

. . . regarding abnormal chip ratios, a request was made to obtain historical 

computerized data containing chip readings from 1981 to the present. These 

were to be viewed to determine if certain worker groups had higher rates of 

abnormal ratios than other workers groups. This would provide additional 

information about the existence and relative degree of past neutron exposures to 

the workforce. However, problems were encountered in retrieving the data. 

First, the storage format and type of dosimetry reading equipment changed 

through time and would have required substantial time and resources to recover. 

Secondly, the older dosimeter equipment (Harshaw 2276) was a direct-reading 

machine and did not create a date field when the results were stored. Finally, 

many archive tapes had been reused, and most of the historical data was 

overwritten with more recent data. Consequently, data available for this purpose 

was limited to only the most recent measurements (1992–1995). Thus, it was not 

feasible to reconstruct neutron exposures from previous TLD chip readings 

before 1992. [Emphasis added.] 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

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In summary, NIOSH’s HHE of the PORTS facility concluded the following: 

•	 The observed anomalous chip ratios were not the result of defective TLDs, but of neutron 

exposures to slow cookers 

•	 Uranium deposits that produced slow cookers existed since the startup of PORTS 

•	 Although TLD badges used since 1981 had the potential for assessing neutron exposures, 

the dosimeters were neither calibrated nor processed to assess neutron dose 

•	 For the time period of 1981 through 1992, historical dosimetry records of anomalous chip 

ratios are either irretrievable or no longer exist 

Although the current technical basis document (ORAUT-TKBS-0015-6) acknowledges the HHE 

findings (and recommends use of the study’s area neutron dose rates and neutron-to-photon 

ratios for dose reconstruction), the TBD ignores the issue of past unaccounted neutron doses 

associated with slow cookers. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

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factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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5.0	 WORKER INTERVIEWS, DATA COMPLETENESS, AND DATA 

INTEGRITY 

The PORTS Site Profile does not adequately address outstanding concerns that have been raised 

regarding the completeness and integrity of dosimetry records representing time periods within 

and outside the SEC. 

5.1 RELEVANT BACKGROUND INFORMATION 

In addressing questions relating to the credibility and adequacy of monitoring data for use in 

dose reconstruction, it is important to recognize that data quality and data completeness are 

subjective terms. The PORTS facility has operated for more than 50 years, during which time 

industry-wide standards, monitoring practices, and technologies have steadily 

changed/improved. 

Regulations defined in behalf of EEOICPA fully anticipated some limitations and deficiencies 

regarding past methods for monitoring workers, the availability of monitoring records, etc. To 

compensate for these deficiencies, regulatory guidance defined in 42 CFR 82 specifies the use of 

surrogate data in dose reconstruction on the condition that such data is technically defensible and 

claimant favorable. 

The ability to reconstruct doses may, nevertheless, be challenged if select conditions exist that 

are defined in §83.9 of 42 CFR 83. Conditions that may preclude a credible dose reconstruction 

are the following: 

(i) Documentation or statements provided by affidavit indicating that 

radiation exposures and doses to members of the proposed class were not 

monitored, either through personal or area monitoring; or 

(ii) Documentation or statements provided by affidavit indicating that 

radiation monitoring records for members of the proposed class have been lost, 

falsified, or destroyed; or 

(iii) A report from a health physicist or other individual with expertise in 

dose reconstruction documenting the limitations of existing DOE or AWE records 

on radiation exposures at the 

facility, as relevant to the petition. This report should specify the basis for 

believing these documented limitations might prevent the completion of dose 

reconstructions for members of the class under 42 CFR Part 82 and related 

NIOSH technical implementation guidelines; or 

(iv) A scientific or technical report, published or issued by a government 

agency of the Executive Branch of government or the General Accounting Office, 

the Nuclear Regulatory Commission, or the Defense Nuclear Facilities Safety 

Board, or published in a peer-reviewed journal, that identifies dosimetry and 

related information that are unavailable (due to either a lack of monitoring or the 

destruction or loss of records) for estimating the radiation doses of employees 

covered by the petition. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

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To determine if any of the above-cited conditions existed to a level that would preclude the dose 

reconstruction for non-presumptive cancers, SC&A conducted interviews with PORTS 

workers/site experts and reviewed a variety of relevant documents/reports. 

5.2 SITE INTERVIEWS 

An integral part of SC&A’s review of any site profile is to conduct interviews with site expert(s) 

and site workers/union representatives. Thus, during a site visit to the PORTS facility on May 

20–26, 2007, several current and former employees expressed concerns regarding the quality, 

completeness, and potential corruption of personnel monitoring records. These very concerns 

have previously been raised by others, including members of Local #66 of the Security, Police, 

and Fire Professionals of America (SPFPA) Union and members of the United Steel Workers 

(USW) Union. SC&A was told that these allegations had been shared with the Department of 

Labor in a letter dated July 12, 2005 (Bowe 2005), as well as with NIOSH during a worker 

outreach meeting between SPFPA and NIOSH in Piketon, Ohio, on November 30, 2005. At the 

time of the worker outreach meeting, NIOSH was provided with copies of two documents that 

addressed these allegations in detail (i.e., the Butler 1996 Report and the Cardarelli 1997 

Report), and was asked to review an evaluation of the PORTS facility by the DOE (DOE 2000). 

In summary, members of the SPFPA and USW had requested NIOSH to review these documents 

in context with their allegations and assess their potential impact on the feasibility of worker 

dose reconstruction. 

Most relevant to the feasibility of dose reconstruction among allegations raised by PORTS 

workers are the following: 

(1) Generic Concerns 

•	 Workers with long-term employment witnessed the many changes/improvements in 

worker protection and worker monitoring that only occurred in the more recent years. 

Generic concern among these workers was that during peak production periods, 

worker protection was not a top priority; health physics personnel were grossly 

understaffed and lacked experience; and engineering controls were inadequate. Thus, 

the time periods with the highest potential for worker exposures coincided with the 

lowest standards for protecting and monitoring workers. 

•	 Radiological incidents involving spills, fires, gaseous leaks were many and routine. 

However, the threshold for documenting such incidents was arbitrary. For example, 

equipment failures such as those involving pigtails to cylinders in which UO2F2 was 

released were not always documented/investigated and not all workers who may have 

been exposed were monitored. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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(2) Concerns for Select Workers 

Among the worker groups interviewed, special concerns were voiced by security 

personnel. For obvious reasons, site security was an integral part of all PORTS activities. 

Thus, security personnel were not only granted unfettered access to all buildings that 

processed radioactive materials, but were also responsible for around-the-clock security 

of SNM, LEU, and HEU during onsite transit and storage. The latter assignments placed 

security personnel in close proximity to these sources for lengthy periods of time. 

Unique to security personnel was the absence or reduced radiological protection afforded 

to them with regard to worker training in radiological protection, use of anticontamination clothing, use of respirators, and routine bioassay sampling during the 

1970s and 1980s. 

(3) Concerns Pertaining to Unmonitored Neutron Exposures 

•	 Since the introduction of TLDs at PORTS in 1981, “abnormal” chip ratios have 

been routinely observed, but were never investigated by health physics personnel 

who dismissed them as artifacts of defective TLDs 

•	 The “abnormal” chip ratios are suggestive of significant exposure to neutrons 

•	 However, prior to 1992, the personnel TLDs were neither calibrated for neutron 

exposures nor processed for potential neutron exposures 

•	 Dosimetry records for this time period that might have documented these 

abnormal chip ratios, however, are no longer available 

(4) Concerns Regarding Dosimeter Data and the Integrity of Records 

•	 Personnel dosimeters were frequently processed in behalf of a “generic bar code” 

that subsequently could not be linked to the individual to whom the dosimeter had 

been assigned. These dosimeters were referred to as the “bucket dosimeters.” 

•	 Documented allegations suggest that official dosimetry records were altered. In 

one specific case, the recorded shallow and deep dose of 26 mrem was replaced 

with a dose of 0, but was subsequently restored to their original value of 26 mrem. 

A comprehensive assessment of workers’ concerns is enclosed herein as Appendix 1. 

Appendix 1 represents comments provided during interviews of 46 current and former workers at 

PORTS whose employment periods range from 1954 to the present. The experience among 

those interviewed range from production, laboratory support, facility maintenance, 

environmental monitoring, medical security, and health physics. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

 

 

 

 

 

 

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5.3 DOCUMENT REVIEW 

In response to concerns raised during the interview sessions, SC&A carefully assessed the 

following documents: 

(1) Memorandum from Don Butler (Lockheed Martin) to Dan Hupp (Lockheed Martin). 

Internal Investigation into Health Physics Management Practices, POEF-150-96-0088, 

February 16, 1996. 

(2) Cardarelli, J., 1996. HETA 96-0198-2651, Portsmouth Gaseous Diffusion Plant, 

Piketon, Ohio, National Institute for Occupational Safety and Health, Cincinnati, Ohio. 

(3) DOE (Department of Energy) 2000. Independent Investigation of the Portsmouth 

Gaseous Diffusion Plant – Volume 1: Past Environment, Safety, and Health Practices, 

DOE Office of Oversight, Environment, Safety and Health. 

(4) Public comments provided by SPFPA at the June 14, 2006, Advisory Board Meeting in 

Washington, DC (ABRWH 2006). 

(5) Letter from J.W. Neton (NIOSH) to D. Bowe (President, SPFPA Local #66), dated 

December 19, 2005, Subject: Discussion of Portsmouth’s External Dosimetry Program. 

National Institute for Occupational Safety and Health, Office of Compensation Analysis 

and Support. 

(6) E-mail correspondence from J.W. Neton (NIOSH) to D.C. Bowe (President, SPFPA 

Local #66), dated December 16, 2005, RE: Discussion, National Institute for 

Occupational Safety and Health, Office of Compensation Analysis and Support. 

(7) Notes prepared by SPFPA regarding dosimetry practices and worker safety. 

Of primary importance to the issue of data integrity are the first three documents. Presented 

below are summary statements, background information, and tentative conclusions cited in each 

of the reports. 

5.4 THE BUTLER 1996 REPORT 

This 41-page report describes an internal investigation of the PORTS health physics practices 

that resulted from specific allegations of improper conduct by a member of the health physics 

staff. The investigation was performed by a three-member team. Allegations relevant to data 

integrity included the following: 

(1) The improper changing of the recorded dose of 26 mrem assigned to a worker (who was 

a security guard at PORTS) to a 0 dose. The 0 dose was subsequently changed back to 

the previously recorded dose of 26 mrem. 

(2) The corruption in the maintenance of the DOELAP TLD database. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

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(3) Misrepresentations of personnel training records to outside auditors in behalf of the 

National Voluntary Laboratory Accreditation Program (NVLAP). 

In response to these allegations, the investigation team from Lockheed Martin offered the 

following opinions and conclusions: 

•	 On the allegations of improper changing of recorded dose: 

It is difficult to understand why such an improper action would have been 

taken upon the negligible figure of 26 millirem. If true, such an action 

represents a case of gross impropriety on the part of [name deleted]. The fact 

that the figures were subsequently changed back to the previous format 

further complicates the issue, as it raised the possibility of “cold feet” for the 

individual responsible for the change. 

Although we believe the allegation to be true, we cannot definitively prove it. 

•	 On the allegation of corrupting the TLD database: 

[Names withheld] describe the DOELAP TLD database as being basically 

reliable, erring on the conservative side, if anything. While a number of 

unassignables obviously exist, several experienced Health Physics personnel 

believe that the data is quite sound ([Name withheld] and [Name withheld], 

however, maintained strong reservations about the validity of the DOELAP 

TLD Database). 

The function of this database is now relegated to that of a historical file. 

Thus, while this allegation does contain some truth, the original assessment of 

large numbers of misassignments has, to say the least, not been proven. . . . 

In conclusion, the absolute validity of the DOELAP TLD Database cannot, at 

this time, be verified. 

•	 On the allegation that health physics supervision misrepresented training records/ 

qualifications of a technician during a NVLAP audit: 

By [name withheld] own admission, [name withheld] was not fully qualified to 

process dosimeters. It seems probable, however, that the auditor was 

persuaded otherwise, either by [name withheld] or by [name withheld]. The 

auditor could have, however, simply been shown the limited training records, 

and allowed to draw his own conclusions. [Name withheld] stated that he 

could understand why they passed the audit in regards to training, because, 

after all, no TLDS were yet being processed through the system, and the 

auditor appeared to be pleased with the planned training program. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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This allegation thus appears to have, at least, a limited amount of support, but 

is neither confirmed or denied. 

5.5 THE CARDARELLI (1997) REPORT 

On June 12, 1996, NIOSH received a request for a Health Hazard Evaluation (HHE) that was to 

focus on worker exposures to neutrons at PORTS. In response to this request, NIOSH conducted 

two site visits between November 1996 and February 1997 (Cardarelli 1997), with the following 

six objectives: 

(1) Determine if potential neutron exposures exist at the site 

(2) Identify neutron sources 

(3) Identify work areas or job titles having the greatest potential for neutron exposures 

(4) Quantify neutron doses by work area or job title 

(5) Determine past reporting and recording practices regarding neutron doses 

(6) Assess the feasibility of reconstructing past neutron doses 

The HHE clearly showed that (1) neutron radiation has been present over the entire period of the 

PORTS facility operation, (2) significant neutron radiation sources are/were present in areas 

where uranium compounds are present in appreciable amounts, and (3) at the time of the HHE, 

neutron dose rates varied, based on work location and job function. 

With regard to objectives #5 and #6, the HHE stated the following (Cardarelli 1997): 

... personnel neutron dosimetry was not conducted . . . in the past . . . 

The early film badges (1950s–1980) and the first TLD badges (1981–1990) were 

not calibrated and could not measure neutron exposures. Despite that limitation, 

an effort to reconstruct past neutron exposures was attempted by requesting 

historical computerized data of TLD chip readings from 1981 to the present. 

Data available for this purpose was limited to only the most recent data (1992– 

1995). Thus it was not feasible to reconstruct potential neutron [doses] before 

1992. 

5.6 DOE (2000) REPORT 

A significant portion of this report assessed the quality of worker health and safety programs at 

PORTS, and documented program deficiencies, worker allegations of unsafe conditions, and 

management’s relationship to union workers. The following statements cited in the DOE (2000) 

not only summarize noted deficiencies and allegations, but also reveal circumstances that may 

impact their interpretation. 

• On Programmatic Deficiencies: 

...appraisal of ES&H [Environmental Safety and Health], called “contractor 

health protection program reviewers,” were performed as early as 1957, and the 

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AEC manual required annual ES&H assessments starting in 1961. These 

assessments were generally performed by two persons over three days and 

addressed radiation protection, criticality safety, industrial hygiene, 

environmental programs... Although important deficiencies and issues were 

identified by these reviews, the size and complexity of Plant operations and the 

nature of the industrial hazards... warranted longer and more frequent 

assessments. 

...A more in-depth, two-week assessment conducted in 1973 by OR [DOE Oak 

Ridge Operations Office] included field observations of Plant conditions, work 

performance, and interviews with workers and first-line supervisors; it concluded 

that the health protection program at PORTS was inadequate. However, there 

was no further evidence of more rigorous assessments, and the limited annual 

appraisals resumed until the 1980s. 

The AEC and its successor organizations also investigated worker allegations of 

unsafe conditions and practices, but with inconsistent rigor and results. ... From 

1979 through 1982, another major DOE investigation of worker complaints, 

conducted at the direction of Congress, identified performance problems in a 

variety of ES&H areas. 

The 1989 DOE Tiger Team assessment identified numerous health, safety, and 

environmental deficiencies... 

Historical weaknesses in DOE investigations of workers allegations have 

continued to the present program. One case in particular, raised during the 

transition from DOE to NRC oversight of USEC, still remains unresolved. That 

case involves allegations by a Plant guard who maintains that his radiation 

exposure records were falsified. Internal investigations by Lockheed Martin 

Utility Services found some merit to the allegation [Note: this is a reference to 

the Butler 1996 Report], and the allegation was forwarded to the Oak Ridge 

Operations Office Inspector General in 1996. That case remains inadequately 

investigated and unresolved by DOE. 

• On Labor Relations (pp. 77–80): 

Established in 1954, the Oil, Chemical, and Atomic Workers Union (OCAW) was 

aggressive in its efforts to protect and improve employee welfare. This 

aggressiveness sometimes caused friction between Plant management and labor. 

On numerous occasions, the positions of management and labor differed widely, 

and resolution was accompanied by extreme measures, as evidenced by one 

unauthorized and six authorized strikes that occurred from 1954 to 1993. 

Furthermore, the severity of management and labor disagreements appears to 

have increased beginning in 1974, as suggested by the frequency and duration of 

strikes. While economic issues were common to most strikes, safety and health 

were an important element in three of these seven actions... 

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Collectively, the number of grievances filed, worker compensation claims 

submitted, and alleged acts of retaliation committed provide further support that 

management and labor relations were strained. From 1954 through 1993, it is 

estimated that more than 17,000 union worker grievances were filed addressing a 

variety of issues in addition to safety and health, including work jurisdiction, 

discipline, overtime, work rules, and benefits.. ... 

... there are records suggesting that labor grievances were filed to be 

confrontational, management appeared to have been acting appropriately and in 

the interest of its employees’ safety and health. 

From pp. 5-6: 

The other major union at the Plant, the United Plant Guard Workers of America 

(UPGWA), has had no strikes since its formation in 1955. ... 

Relations between... UPGWA union and Plant management were much less 

confrontational. Although protective forces have been an integral part of Plant 

activities due to security considerations, the ES&H protection provided to 

production workers (such as respirators and shoe covers) were not always 

considered or provided to security personnel when they worked in close proximity 

to hazardous operations or were stationed, ate lunch, and took breaks in 

contaminated areas. In addition, in the late 1980s and early 1990s, protective 

forces performed extensive training drills in radioactively contaminated buildings 

without appropriate protective clothing or monitoring. Hazard communications 

and ES&H training have not always been provided on a timely and consistent 

basis for protective force workers. 

5.7	 CONCLUSIONS PERTAINING TO DATA INTEGRITY AND THE ADEQUACY 

OF DATA FROM DOSE RECONSTRUCTION 

On the basis of available information gained from the review of the PORTS Site Profile, site 

expert interviews, and the previously cited documents, SC&A concludes the following: 

• On programmatic deficiencies and gaps in personnel monitoring: 

Noted deficiencies in monitoring personnel for internal and external exposures are real, 

but are not unique to PORTS workers. To date, NIOSH and its contractors have 

encountered similar deficiencies at other facilities that have been resolved by various 

means (i.e., use of surrogate data, coworker models, etc.). 

• On the falsification of records: 

SC&A agrees that the falsification of dosimetry data is a serious matter that should be 

thoroughly investigated. Documented evidence of altering personnel doses at PORTS, 

however, appears to be restricted to a single individual and for a single dosimeter wear 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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period. Moreover, the initial dose of 26 mrem that was changed to a 0 dose was 

subsequently restored to its original value. 

Although no explanation has been offered that would justify the described events, SC&A 

concludes that this isolated event does not adversely impact the reconstruction of worker 

doses at PORTS. 

• On the “corruption” of the DOELAP TLD Database: 

SC&A believes that persistent perceptions regarding data integrity issues have been 

exacerbated by (1) personality conflicts among select individuals within the health 

physics group, and (2) a history of poor relations and mistrust between labor and 

management at PORTS. 

Over the years, many improvements were introduced in personnel monitoring. Key 

changes included period replacement of dosimetry systems that started with manually 

processed film dosimeters and evolved to increasingly automated TLD systems. The 

transition from one system to the next undoubtedly may have introduced problems, such 

as the alleged problem associated with the DOELAP dosimeters’ bar codes. 

SC&A concludes that while the bar code problem with the DOELAP TLD may have 

resulted in some unassignable doses, the likely number of affected persons were small. 

More importantly, any discrepancy between recorded and actual doses does not reflect a 

deliberate attempt by health physics personnel to reduce/falsify worker exposures. 

• On unmonitored neutron doses: 

Based on the physical nature and material quantities processed at PORTS, neutron 

exposures should have been anticipated for the entire period of facility operation. 

Therefore, the failure to properly monitor PORTS workers for neutron exposure prior to 

1992 must be regarded as a deficiency that has to be accounted for in dose reconstruction. 

ORAUT-TKBS-0015-6 references multiple studies inclusive of the 1997 Cardarelli 

study. In combination, these studies provide surrogate data that may reasonably reflect 

routine unmonitored neutron exposures. A major limitation of the surrogate neutron 

dose models is their failure to account for “unusual” neutron exposures, as documented 

by the observed abnormal chip ratios of the DOELAP TLDs. This limitation was 

acknowledged by Cardarelli (1997), who concluded that “. . . it was not feasible to 

reconstruct neutron exposures from previous [abnormal] TLD chip readings before 

1992.” 

Currently, the unresolved failure to account for neutron exposures associated with 

abnormal chip ratios that may reflect “slow cookers” is discussed as Finding 4.6-4 in this 

report and must be resolved. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

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6.0 REFERENCES 

42 CFR Part 82. Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule, Federal Register/Vol. 67, 

No. 85/Thursday, May 2, 2002, p 22314. 

1957 Committee Report, 1957. Argon Gamma Graph Committee and Operations Specifications, 

FR-434-06-0035, Box 11. 

Bassett, A.C., 1986. Thermoluminescent Dosimeter Response and Calibration, GAT-S-56, 

Piketon, Ohio, April 20. 

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NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

Effective Date: 

February 6, 2008 

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NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

Effective Date: 

February 6, 2008 

Revision No. 

0 

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NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

 

Effective Date: 

February 6, 2008 

Revision No. 

0 

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SC&A 2007. Review of the Feed Materials Production Center (FMPC) Special Exposure 

Cohort (SEC) Petition-00046 and the NIOSH SEC Petition Evaluation Report, Draft Report. S. 

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and Health Physics, Goodyear Atomic Corporation, Portsmouth, Ohio, June 1, 1964. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.
 
 

 

Effective Date: 

February 6, 2008 

Revision No. 

0 

Document No. 

SCA-TR-TASK1-0020 

Page No. 

89 of 89 

ATTACHMENT 1: SUMMARY OF SITE EXPERT INTERVIEWS 

This attachment will be provided at a later date. 

NOTICE: This report (with the exception of Attachment 1, which is not included here) has been reviewed for Privacy Act 

information and has been cleared for distribution on February 6, 2008. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker Health for 

factual accuracy or applicability within the requirements of 42 CFR 82.